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Sample records for gamma skyshine calculations

  1. The integral first collision kernel method for gamma-ray skyshine analysis[Skyshine; Gamma-ray; First collision kernel; Monte Carlo calculation

    Energy Technology Data Exchange (ETDEWEB)

    Sheu, R.-D.; Chui, C.-S.; Jiang, S.-H. E-mail: shjiang@mx.nthu.edu.tw

    2003-12-01

    A simplified method, based on the integral of the first collision kernel, is presented for performing gamma-ray skyshine calculations for the collimated sources. The first collision kernels were calculated in air for a reference air density by use of the EGS4 Monte Carlo code. These kernels can be applied to other air densities by applying density corrections. The integral first collision kernel (IFCK) method has been used to calculate two of the ANSI/ANS skyshine benchmark problems and the results were compared with a number of other commonly used codes. Our results were generally in good agreement with others but only spend a small fraction of the computation time required by the Monte Carlo calculations. The scheme of the IFCK method for dealing with lots of source collimation geometry is also presented in this study.

  2. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  3. GRAYSKY-A new gamma-ray skyshine code

    International Nuclear Information System (INIS)

    Witts, D.J.; Twardowski, T.; Watmough, M.H.

    1993-01-01

    This paper describes a new prototype gamma-ray skyshine code GRAYSKY (Gamma-RAY SKYshine) that has been developed at BNFL, as part of an industrially based master of science course, to overcome the problems encountered with SKYSHINEII and RANKERN. GRAYSKY is a point kernel code based on the use of a skyshine response function. The scattering within source or shield materials is accounted for by the use of buildup factors. This is an approximate method of solution but one that has been shown to produce results that are acceptable for dose rate predictions on operating plants. The novel features of GRAYSKY are as follows: 1. The code is fully integrated with a semianalytical point kernel shielding code, currently under development at BNFL, which offers powerful solid-body modeling capabilities. 2. The geometry modeling also allows the skyshine response function to be used in a manner that accounts for the shielding of air-scattered radiation. 3. Skyshine buildup factors calculated using the skyshine response function have been used as well as dose buildup factors

  4. Evaluation of skyshine dose due to gamma-rays from a cobalt-60 irradiation facility

    International Nuclear Information System (INIS)

    Kanazawa, Tamotsu; Okamoto, Shinichi; Ohnishi, Tokuhiro; Tsujii, Yukio

    1991-01-01

    We attempted to evaluate skyshine dose due to gamma-rays from a cobalt-60 irradiation facility. As the first step, the results of measurements and calculations were compared of the skyshine dose due to gamma-rays from the cobalt-60 source of 1.45 PBq set in the No.4 irradiation room of our laboratory. Distances of measuring points from the cobalt source were in the range from 17 m to about 100 m in the site of our office. Calculation was carried out with simplified single scattering method. The calculated values of the skyshine dose were higher than the measured values. For more precise evaluation of the skyshine dose, the following factors are to be considered; the dose rate distribution on the roof above the source and the attenuation of gamma-rays by air. (author)

  5. Computational techniques in gamma-ray skyshine analysis

    International Nuclear Information System (INIS)

    George, D.L.

    1988-12-01

    Two computer codes were developed to analyze gamma-ray skyshine, the scattering of gamma photons by air molecules. A review of previous gamma-ray skyshine studies discusses several Monte Carlo codes, programs using a single-scatter model, and the MicroSkyshine program for microcomputers. A benchmark gamma-ray skyshine experiment performed at Kansas State University is also described. A single-scatter numerical model was presented which traces photons from the source to their first scatter, then applies a buildup factor along a direct path from the scattering point to a detector. The FORTRAN code SKY, developed with this model before the present study, was modified to use Gauss quadrature, recent photon attenuation data and a more accurate buildup approximation. The resulting code, SILOGP, computes response from a point photon source on the axis of a silo, with and without concrete shielding over the opening. Another program, WALLGP, was developed using the same model to compute response from a point gamma source behind a perfectly absorbing wall, with and without shielding overhead. 29 refs., 48 figs., 13 tabs

  6. Recent skyshine calculations at Jefferson Lab

    International Nuclear Information System (INIS)

    Degtyarenko, P.

    1997-01-01

    New calculations of the skyshine dose distribution of neutrons and secondary photons have been performed at Jefferson Lab using the Monte Carlo method. The dose dependence on neutron energy, distance to the neutron source, polar angle of a source neutron, and azimuthal angle between the observation point and the momentum direction of a source neutron have been studied. The azimuthally asymmetric term in the skyshine dose distribution is shown to be important in the dose calculations around high-energy accelerator facilities. A parameterization formula and corresponding computer code have been developed which can be used for detailed calculations of the skyshine dose maps

  7. Secondary gamma-ray skyshine from 14 MeV Neutron Source Facility (OKTAVIAN). Comparison of measurement with its simulation

    Energy Technology Data Exchange (ETDEWEB)

    Morotomi, Ryutaro; Kondo, Tetsuo; Murata, Isao; Yoshida, Shigeo; Takahashi, Akito [Osaka Univ., Department of Nuclear Engineering, Suita, Osaka (Japan); Yamamoto, Takayoshi [Osaka Univ., Radio Isotope Research Center, Suita, Osaka (Japan)

    2000-03-01

    Measurement of secondary gamma-ray skyshine was performed at the Intense 14 MeV Neutron Source Facility (OKTAVIAN) of Osaka University with NaI and Hp-Ge detectors. From the result of measurements, some mechanism of secondary gamma-ray skyshine from 14 MeV neutron source facility was found out. The analysis of the measured result were carried out with MCNP-4B for four nuclear data files of JENDL-3.2, JENDL-F.F., FENDL-2, and ENDF/B-VI. It was confirmed that all the nuclear data are fairly reliable for calculations of secondary gamma-ray skyshine. (author)

  8. Skyshine spectra of gamma rays

    International Nuclear Information System (INIS)

    Swarup, Janardan

    1980-01-01

    A study of the spectra of gamma photons back-scattered in vertical direction by infinite air above ground (skyshine) is presented. The source for these measurements is a 650 Ci Cobalt-60 point-source and the skyshine spectra are reported for distances from 150 m to 325 m from the source, measured with a 5 cm x 5 cm NaI(Tl) detector collimated with collimators of 12 mm and 20 mm diameter and 5 cm length. These continuous spectra are unfolded with Gold's iterative technique. The photon-spectra so obtained have a distinct line at 72 keV due to multiply-scattered photons. This is an energy where photoelectric and Compton cross-sections for multiply-scattered photons balance each other. The intensity of the line(I) decreases exponentially with distance (d) from the source obeying a relation of the type I = Isub(o)esup(-μd) where μ is called as ''Multiply-Scatter Coefficient'', a constant of the medium which is air in these measurements. This relationship is explained in terms of a halo around the source comprising of multiply-scattered gamma photons, Isub(0) being the intensity of these scattered photons at the location of cobalt-source. A fraction called as ''Back-scattered Fraction'', the ratio of Isub(0) to the number of original photons from the cobalt-source entering the infinite air, is also calculated. It is shown that with a properly calibrated detector system, this fraction can be used to determine the strength of a large gamma source, viz. a nuclear explosion in air, and for mineral prospecting. These conclusions are general and can be applied to any other infinite medium. Some forward-scatter (transmission) spectra of cobalt-60 source through 10 cm of Pb and 2.5 cm of Al are also reported. (auth.)

  9. Measurement of secondary gamma-ray skyshine and groundshine from intense 14 MeV neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Shigeo; Morotomi, Ryutaro; Kondo, Tetsuo; Murata, Isao; Takahashi, Akito [Osaka Univ., Suita (Japan). Dept. of Nuclear Engineering

    2000-03-01

    Secondary gamma-ray skyshine and groundshine, including the direct contribution from the facility building, have been measured with an Hp-Ge detector and an NaI(Tl) detector at the Intense 14 MeV Neutron Source Facility OKTAVIAN of Osaka University, Japan. The mechanism of secondary gamma-rays propagation were analyzed with the measured spectrum with the Hp-Ge detector. The contribution of the skyshine was shown to be a continuum spectrum that was composed of mainly Compton scattered high energy secondary gamma-rays generated in the facility building created by (n, {gamma}) reaction. The contribution of the groundshine considerably contained secondary gamma-rays generated by {sup nat}Si (n, {gamma}) reaction in soil, including the albedo contribution from the ground. And the total contribution contained capture gamma-rays from iron (Fe) and other nuclides. The measurements with the NaI(Tl) detector as well as the Hp-Ge detector were carried out to investigate the dependence of gamma-ray dose as a function of distance from the neutron source up to hundreds meters. Consequently, it was found that the dependence could be fitted with the function of const.{center_dot}exp(-r/{lambda})/r{sup n}, where n values were around 2 except for the skyshine (n {approx} 1). It was thus indicated that the contribution of the skyshine could be propagated farther downfield than the direct contribution from the facility. The measured ratios of the three contributions (skyshine, groundshine, and direct contributions) and the distance dependence in each path were shown to be in good agreement with calculated results by the Monte Carlo transport code MCNP-4A. And the total contributions for the two detectors of NaI(Tl) and Hp-Ge agree excellently with each other. (author)

  10. Monte Carlo calculation of ''skyshine'' neutron dose from ALS [Advanced Light Source

    International Nuclear Information System (INIS)

    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

  11. Improvement of air transport data and wall transmission/reflection data in the SKYSHINE code. 2. Calculation of gamma-ray wall transmission and reflection data

    Energy Technology Data Exchange (ETDEWEB)

    Hayashida, Yoshihisa [Toshiba Corp., Kawasaki, Kanagawa (Japan); Ishikawa, Satoshi; Harima, Yoshiko [CRC Research Institute Inc., Tokyo (Japan); Hayashi, Katsumi; Tayama, Ryuichi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nemoto, Makoto [Visible Information Center, Tokai, Ibaraki (Japan); Sato, Osamu [Mitsubishi Research Inst., Inc., Tokyo (Japan)

    2000-03-01

    Transmission and reflection data of concrete and steel for 6.2 MeV gamma-ray in the SKYSHINE code have been generated using up-to-date data and method with a view to improving an accuracy of results. The transmission and reflection data depend on energy and angle. The invariant embedding method, which has merits of producing no negative angular flux and of taking small computer time, is suitable and adopted to the present purpose. Transmission data were calculated for concrete of 12 {approx} 160 cm thick and steel of 4 {approx} 39 cm thick based on the PHOTX library. Reflection data were calculated for semi-infinite slabs of concrete and steel. Consequently, smooth and consistent differential data over whole angle and energy were obtained compared with the original data calculated by discrete ordinates Sn code and Monte Carlo code. In order to use these data in the SKYSHINE code, further verification is needed using various calculation method or experimental data. (author)

  12. Applications of Monte Carlo codes to a study of gamma-ray buildup factors, skyshine and duct streaming

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    Hirayama, H. [High Energy Accelerator Research Organization (KEK), Ibaraki (Japan)

    2001-07-01

    Many shielding calculations for gamma-rays have continued to rely on point-kernel methods incorporating buildup factor data. Line beam or conical beam response functions, which are calculated using a Monte Carlo code, for skyshine problems are useful to estimate the skyshine dose from various facilities. A simple calculation method for duct streaming was proposed using the parameters calculated by the Monte Carlo code. It is therefore important to study, improve and produce basic parameters related to old, but still important, problems in the fields of radiation shielding using the Monte Carlo code. In this paper, these studies performed by several groups in Japan as applications of the Monte Carlo method are discussed. (orig.)

  13. Effect of source angular distribution on the evaluation of gamma-ray skyshine

    Energy Technology Data Exchange (ETDEWEB)

    Sheu, R.D.; Jiang, S.H. [Dept. of Engineering and System Science, National Tsing Hua Univ., Taiwan (China); Chang, B.J.; Chen, I.J. [Division of Health Physics, Inst. of Nuclear Energy Research, Taiwan (China)

    2000-03-01

    The effect of the angular distribution of the equivalent point source on the analysis of the skyshine dose rates was investigated in detail. The dedicated skyshine codes SKYDOSE and McSKY were revised to include the capability of dealing with the anisotropic source. It was found that a replace of the cosine-distributed source by an isotropic source will overestimate the skyshine dose rates for large roof-subtended angles and cause underestimation for small roof-subtended angles. For building with roof shielding, however, replacing the cosine-distributed source by an isotropic source will always underestimate the skyshine dose rates. The skyshine dose rates from a volume source calculated by the dedicated skyshine code agree very well with those of the MCNP Monte Carlo calculation. (author)

  14. Monte Carlo method for calculating the radiation skyshine produced by electron accelerators

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    Kong Chaocheng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China)]. E-mail: kongchaocheng@tsinghua.org.cn; Li Quanfeng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Chen Huaibi [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Du Taibin [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Cheng Cheng [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Tang Chuanxiang [Department of Engineering Physics, Tsinghua University Beijing 100084 (China); Zhu Li [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Zhang Hui [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Pei Zhigang [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China); Ming Shenjin [Laboratory of Radiation and Environmental Protection, Tsinghua University, Beijing 100084 (China)

    2005-06-01

    Using the MCNP4C Monte Carlo code, the X-ray skyshine produced by 9 MeV, 15 MeV and 21 MeV electron linear accelerators were calculated respectively with a new two-step method combined with the split and roulette variance reduction technique. Results of the Monte Carlo simulation, the empirical formulas used for skyshine calculation and the dose measurements were analyzed and compared. In conclusion, the skyshine dose measurements agreed reasonably with the results computed by the Monte Carlo method, but deviated from computational results given by empirical formulas. The effect on skyshine dose caused by different structures of accelerator head is also discussed in this paper.

  15. Calculation of dose equivalents for photon skyshine production; Calculo da dose equivalente para fotons decorrente da producao de skyshine

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    Frota, Marco A.; Kelecom, Alphonse [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Dept. de Biologia Geral. Lab. de Radiobiologia e Radiometria (LARARA)]. E-mail: egbakel@vm.uff.br

    2005-07-01

    Some radiation facilities are designed with little shielding in the ceiling above the accelerator. A problem may then arise as a result of the radiation scattered by the atmosphere to points at ground level outside the treatment room. Stray radiation of this type is referred to as skyshine, and the National Council on Radiation Protection and Measurements Report No. 51 (NCRP 1977) gives methods for the calculation of skyshine for accelerator facilities. McGinley (1993) has compared skyshine measurements made at an 18 MeV medical accelerator facility with values calculated using the techniques presented in NCRP Report No. 51. Measurements were made of the photon levels outside a treatment room housing a Varian 2100 deg C. The roof above the accelerator was designed for weather protection only and offered little shielding for the primary beam and scattered radiation. The distance from the treatment room floor to the roof was 4.27 m, and the primary walls were constructed of concrete 2.0 m thick.The secondary walls were fabricated of concrete 0.99 m thick. The results for the photon skyshine rate dose as a function of distance from the isocenter using Monte Carlo code, are compared with those in NCRP publication 74 and measured obtained. The photon skyshine dose rates simulated for real clinic spectra transmitted through roof range from 4.7 to 14.6 nSv.s{sup -1}. (author)

  16. Development of 'SKYSHINE-CG' code. A line-beam method code equipped with combinatorial geometry routine

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Takahiro; Ochiai, Katsuharu [Plant and System Planning Department, Toshiba Corporation, Yokohama, Kanagawa (Japan); Uematsu, Mikio; Hayashida, Yoshihisa [Department of Nuclear Engineering, Toshiba Engineering Corporation, Yokohama, Kanagawa (Japan)

    2000-03-01

    A boiling water reactor (BWR) plant has a single loop coolant system, in which main steam generated in the reactor core proceeds directly into turbines. Consequently, radioactive {sup 16}N (6.2 MeV photon emitter) contained in the steam contributes to gamma-ray skyshine dose in the vicinity of the BWR plant. The skyshine dose analysis is generally performed with the line-beam method code SKYSHINE, in which calculational geometry consists of a rectangular turbine building and a set of isotropic point sources corresponding to an actual distribution of {sup 16}N sources. For the purpose of upgrading calculational accuracy, the SKYSHINE-CG code has been developed by incorporating the combinatorial geometry (CG) routine into the SKYSHINE code, so that shielding effect of in-building equipment can be properly considered using a three-dimensional model composed of boxes, cylinders, spheres, etc. Skyshine dose rate around a 500 MWe BWR plant was calculated with both SKYSHINE and SKYSHINE-CG codes, and the calculated results were compared with measured data obtained with a NaI(Tl) scintillation detector. The C/E values for SKYSHINE-CG calculation were scattered around 4.0, whereas the ones for SKYSHINE calculation were as large as 6.0. Calculational error was found to be reduced by adopting three-dimensional model based on the combinatorial geometry method. (author)

  17. Neutron skyshine from nuclear facilities

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Hayashi, Katsumi.

    1984-01-01

    The advance in neutron skyshine research and the significance are first described. Then, skyshine calculation methods in 1980s particularly and the skyshine experiment in Japan with various nuclear facilities (reactors, D-T neutron sources, accelerators) are reviewed. In comparison with such experiment usable as bench mark, the skyshine calculation methods (Monte Carlo method, transport calculation method) are evaluated for their accuracy and merits and demerits. The values by Monte Carlo calculation were in agreement within about 30 % with the experimental values. Those by DOT 3.5 calculation were twice as large as the experimental values. Those by PALLAS calculation were in good agreement in dose with the experimental values, but the spectra were considerably different. The values by SKYSHINE-2 were in good agreement with the experimental values, but since the ground effect was ignored, the values may deviate from the experimental ones if it is taken into account. (Mori, K.)

  18. Russia skyshine experiment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tsubosaka, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawabe, Toshiaki [The Japan Research Institute, Ltd., Tokyo (Japan); Ueki, Kohtaro [National Maritime Research Institute, Tokyo (Japan)

    2001-01-01

    Experimental studies of neutron and gamma-radiation skyshine at nuclear reactor are proceeding in cooperation with Russia, Kazakhstan and Japan as a project of international science technology center (ISTC). Fast neutron streaming from the vertical experimental hole of IVG.1M reactor which has a cylindrical core are analyzed by a monte carlo n-particle transport code (MCNP) with variance reduction methods, in which a weight window method and a cell importance method can be selected. Calculation results on radial distribution of fast neutron flux at 100 cm above the reactor is compared with the experimental values. The calculated values of neutron flux by using the cell importance method, however, is very different from the experimental values at close distance of 10 cm from the center. Skyshine analysis of neutron radiation streaming from the reactor are also carried out by the equivalent source model in which a point source and the detectors are located at 10 cm and 1 m above the ground, respectively. The calculated values of total neutron flux distribution are very close to the experimental values. The effects of the air composition on neutron flux calculation are also investigated. (M. Suetake)

  19. Comparison of radiation measurements and calculations of reactor surroundings for skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tsubosaka, A.; Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawabe, T. [Japan Research Institute, Limited, Osaka (Japan); Zharkov, V.P.; Kartashev, I.A.; Netecha, M.E.; Orlov, Y.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2000-03-01

    ISTC Project 'Experimental Studies of Radiation Scattering in the Atmosphere' were conducted using the IVG-1M and RA reactors by RDIPE in collaboration with IAE NNC RK and JAERI during 1996-1998. The radial distributions of fast neutron flux, thermal neutron flux and gamma radiation dose rate were measured above these two reactors at three heights. Neutron spectra above these two reactors and thermal and fast neutron fluxes over the hollow pipe height in the IVG-1M reactor were also measured in order to determine the radiation characteristics for skyshine analysis. For verifying the computer codes the calculations of reactor surroundings were performed using MCNP and DORT/DOT-3.5. The comparisons between the measurements and the calculations show that MCNP and DORT/DOT-3.5 codes can be widely applied to the shielding problems by selecting properly the calculation conditions. (author)

  20. SKYSHIN: A computer code for calculating radiation dose over a barrier

    International Nuclear Information System (INIS)

    Atwood, C.L.; Boland, J.R.; Dickman, P.T.

    1986-11-01

    SKYSHIN is a computer code for calculating the radioactive dose (mrem), when there is a barrier between the point source and the receptor. The two geometrical configurations considered are: the source and receptor separated by a rectangular wall, and the source at the bottom of a cylindrical hole in the ground. Each gamma ray traveling over the barrier is assumed to be scattered at a single point. The dose to a receptor from such paths is numerically integrated for the total dose, with symmetry used to reduce the triple integral to a double integral. The buildup factor used along a straight line through air is based on published data, and extrapolated in a stable way to low energy levels. This buildup factor was validated by comparing calculated and experimental line-of-sight doses. The entire code shows good agreement to limited field data. The code runs on a CDC or on a Vax computer, and could be modified easily for others

  1. Radiation skyshine calculation with MARS15 for the Mu2e Experiment at Fermilab

    International Nuclear Information System (INIS)

    Leveling, A.F.

    2015-01-01

    The Fermilab Antiproton source is to be re-purposed to provide an 8 kW proton beam to the Mu2e experiment by 1/3 integer, slow resonant extraction. Shielding provided by the existing facility must be supplemented with in-tunnel shielding to limit the radiation effective dose rate above the shield in the AP30 service building. In addition to the nominal radiation shield calculations, radiation skyshine calculations were required to ensure compliance with Fermilab Radiological Controls Manual. A complete model of the slow resonant extraction system including magnets, electrostatic septa, magnetic fields, tunnel enclosure with shield, and a nearby exit stairway are included in the model. The skyshine model extends above the beam enclosure surface to 10 km vertically and 5 km radially. (authors)

  2. Improvement of air transport data and wall transmission/reflection data in the SKYSHINE code. (1) Calculation of line beam response function for gamma-ray skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Makoto [Visible Information Center, Tokai, Ibaraki (Japan); Harima, Yoshiko; Ishikawa, Satoshi [CRC Research Inst. Inc., Tokyo (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Hayashi, Katsumi; Tayama, Ryuichi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Hayashida, Yoshihisa [Toshiba Corp., Yokohama, Kanagawa (Japan); Sato, Osamu [Mitsubishi Research Inst., Tokyo (Japan)

    2000-03-01

    The line-beam response functions (LBRFs) of a key component of a computationally simple gamma-ray skyshine analysis are generated using an electron-photon cascade Monte Carlo code EGS4. The LBRFs R(E{sub 0}, {phi}, x) are given with the air-kerma (Gy per photon), 7 photon source energies ranging from 0.5 to 10 MeV, for source-detector distances between 10 and 2,000 meters, and at 19 emission angles from 0 - 170 degrees, as measured from the source-detector axis. Especially, the values of R(E{sub 0}, {phi}=0.0 and 0.1, x) are extremely larger than the ones of LBRFs produced by the point kernel model or the COHORT code. The LBRF is accurately approximated by a four-parameter formula. Values of four parameters for the approximate LBRF are described by monotonic and smooth curves with respect to the energy E{sub 0} and the emitted angle {phi}. (author)

  3. Application of a general purpose user's version of the EGS4 code system to a photon skyshine benchmarking calculation

    International Nuclear Information System (INIS)

    Nojiri, I.; Fukasaku, Y.; Narita, O.

    1994-01-01

    A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)

  4. Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Yuji, Nemoto; Toshihisa, Tsukiyama; Shigeki, Nemezawa [Hitachi. Ltd., Saiwai-cho, Hitachi (Japan); Tadashi, Yamasaki; Hidetsugu, Okada [Chubu Electric Power Company, Inc., Odaka-cho, Midori-ku Nagoya (Japan)

    2007-07-01

    The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)

  5. Public Dose Assessment Modeling from Skyshine by Proton Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Mwambinga, S. A. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Yoo, S. J. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this paper, the skyshine dose by proton accelerator (230 MeV) has been evaluated. The amount of dose by skyshine is related to some influence factors which are emission angle (Height wall), the thickness of ceiling and distance from source to receptor (Human body). Empirical formula is made by using MCNPX code results. It can easily calculate and assess dose from skyshine by proton accelerator. The skyshine doses are calculated with MCNPX code and DCFs in ICRP 116. Thereafter, we made empirical formula which can calculate dose easily and be compared with the results of MCNPX. The maximum exposure point by skyshine is about 5 ∼ 10 m from source. Therefore, the licensee who wants to operate the proton accelerator must keep the appropriate distance from accelerator and set the fence to restrict the approach by the public. And, exposure doses by accelerator depend on operating time and proton beam intensities. Eq. (6) suggested in this study is just considered for mono energy proton accelerator. Therefore, it is necessary to expand the dose calculation to diverse proton energies. Radiations like neutron and photon generated by high energy proton accelerators over 10 MeV, are important exposure sources to be monitored to radiation workers and the public members near the facility. At that case, one of the exposure pathways to the public who are located in near the facility is skyshine. Neutrons and photons can be scattered by the atmosphere near the facility and exposed to public as scattered dose. All of the facilities using high energy radiation and NDI (Non-Destructive Inspection) which is tested at open field, skyshine dose must be taken into consideration. Skyshine dose is not related to the wall thickness of radiation shielding directly.

  6. Spatial and energy distributions of skyshine neutron and gamma radiation from nuclear reactors on the ground-air boundary

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Y.; Netecha, M.E.; Vasiliev, A.P.; Avaev, V.N.; Vasiliev, G.A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Zelensky, D.I.; Istomin, Y.L.; Cherepnin, Y.S. [Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Semipalatinsk-21 (Kazakhstan); Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    A set of measurements on skyshine radiation was conducted at two special research reactors. A broad range of detectors was used in the measurements to record neutron and gamma radiations. Dosimetric and radiometric field measurements of the neutrons and gamma quanta of the radiation scattered in the air were performed at distances of 50 to 1000 m from the reactor during different weather conditions. The neutron spectra in the energy range of 1 eV to 10 MeV and the gamma quanta spectra in the range of 0.1-10 MeV were measured. (author)

  7. Accelerator skyshine: Tyger, tyger, burning bright

    International Nuclear Information System (INIS)

    Stapleton, G.B.; Thomas, R.H.

    1992-06-01

    Neutron skyshine is, in most cases, the dominant source of radiation exposure to the general public from operation of well-shielded, high-energy accelerators. To estimate this exposure, tabulated solutions of the transport of neutrons through the air are frequently used. In previous works on skyshine, these tabular data have been parameterized into simple empirical equations that are easy and fast to use but are limited to distances greater than a few hundred meters from the accelerator. Our current report has refined this earlier work by including more realistic assumptions of neutron differential energy spectrum and angular distribution. These improved calculations essentially endorse the earlier parameterizations but make possible reasonably accurate dose estimates much closer to the skyshine source than before

  8. D-T neutron skyshine experiments at JAERI/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, Takeo; Ochiai, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshida, Shigeo [Tokai Univ., Hiratsuka, Kanagawa (JP)] (and others)

    2003-03-01

    The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of {approx}1.7x10{sup 11} n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9x0.9 m{sup 2} was open during the experimental period. The radiation dose rate outside the target room was measured as far as about 550 m away from the D-T target point with a spherical rem-counter. The highest neutron dose was about 0.5 {mu}Sv/hr at a distance of 30 m from the D-T target point and the dose rate was attenuated to 0.002 {mu}Sv/hr at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 250 m. The neutron spectra were evaluated with a {sup 3}He detector with different thickness of polyethylene neutron moderators. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation detectors. (author)

  9. SHINE-III. Simple code for skyshine dose calculation up to 3 GeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tsukiyama, Toshihisa; Tayama, Ryuichi; Handa, Hiroyuki [Hitachi Engineering Co. Ltd., Ibaraki (Japan)] [and others

    2000-03-01

    Skyshine dose at site boundary is considered as one of the most fundamental issues to get approval of constructing nuclear installations. Skyshine conical beam response functions (CBRF) for high energy neutrons up to 3 GeV are obtained using NMTC-JAERI and MCNP code. This CBRF is fitted to the four parameters equation. Simple code named SHINE-III using this equation with updated data is developed. (author)

  10. Measurement of radiation skyshine with D-T neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, S.; Nishitani, T. E-mail: nisitani@naka.jaeri.go.jp; Ochiai, K.; Kaneko, J.; Hori, J.; Sato, S.; Yamauchi, M.; Tanaka, R.; Nakao, M.; Wada, M.; Wakisaka, M.; Murata, I.; Kutsukake, C.; Tanaka, S.; Sawamura, T.; Takahashi, A

    2003-09-01

    The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of {approx}1.7x10{sup 11} n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9x0.9 m{sup 2} was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D-T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9x10{sup -22} Sv/(source neutron) at a distance of 30 m from the D-T target point and the dose rate was attenuated to 4x10{sup -24} Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m.

  11. Application of improved air transport data and wall transmission/reflection data in the SKYSINE code to typical BWR turbine skyshine

    Energy Technology Data Exchange (ETDEWEB)

    Tayama, Ryuichi; Hayashi, Katsumi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harima, Yoshiko; Ishikawa, Satoshi [CRC Research Institute Inc., Tokyo (Japan); Hayashida, Yoshihisa [Toshiba Corp., Kawasaki, Kanagawa (Japan); Nemoto, Makoto [Visible Information Center, Tokai, Ibaraki (Japan); Sato, Osamu [Mitsubishi Research Inst., Inc., Tokyo (Japan)

    2000-03-01

    Three basic sets of data, i.e. air transport data and material transmission/reflection data, included in the SKYSHINE program have been improved using up-to-data and methods, and applied to skyshine dose calculations for a typical BWR turbine building. The direct and skyshine dose rates with the original SKYSHINE code show good agreements with MCNP Monte-Carlo calculations except for the distances less than 0.1 km. The results for the improved SKYSHINE code also have agreements with the MCNP code within 10-20%. The discrepancy of 10-20% can be due to the improved concrete transmission data at small incident and exit angles. We still improve the three sets of data and investigate with different calculational models to get more accurate results. (author)

  12. Projected Standard on neutron skyshine

    International Nuclear Information System (INIS)

    Westfall, R.M.; Williams, D.S.

    1987-07-01

    Current interest in neutron skyshine arises from the application of dry fuel handling and storage techniques at reactor sites, at the proposed monitored retrievable storage facility and at other facilities being considered as part of the civilian radioactive waste management programs. The chairman of Standards Subcommittee ANS-6, Radiation Protection and Shielding, has requested that a work group be formed to characterize the neutron skyshine problem and, if necessary, prepare a draft Standard. The work group is comprised of representatives of storage cask vendors, architect engineering firms, nuclear utilities, the academic community and staff members of national laboratories and government agencies. The purpose of this presentation summary is to describe the activities of the work group and the scope and contents of the projected Standard, ANS-6.6.2, ''Calculation and Measurement of Direct and Scattered Neutron Radiation from Nuclear Power Operations.'' The specific source under consideration by the work group is an array of dry fuel casks located at a reactor site. However, it is recognized that the scope of the standard should be broad enough to encompass other neutron sources. The Standard will define appropriate methodology for properly characterizing the neutron dose due to skyshine. This dose characterization is necessary, for example, in demonstrating compliance with pertinent regulatory criteria

  13. Consideration of sky-shine radiation effects for the development of Korean regulatory guidance about industrial radiography

    International Nuclear Information System (INIS)

    Yong Ki Chi; Bokyun Seo; Wantae Kim

    2015-01-01

    Although most of the sky-shine radiation levels in industrial radiography are below regulatory limits, sky-shine radiation could make a valuable contribution to the total radiation level near shielding facility with little shielding and open field without shielding. Therefore sky-shine radiation should be thoroughly predicted and supervised with the ALARA principle. In this study, we simulated sky-shine radiation for mobile irradiators using MCNP and newly suggested the equation for calculating sky-shine radiation. Also these results were applied to developing Korean regulatory guidance about industrial radiography and to recommending the requirement of the facility design, controlled or supervised area at work places. (author)

  14. On a new method to compute photon skyshine doses around radiotherapy facilities

    Energy Technology Data Exchange (ETDEWEB)

    Falcao, R.; Facure, A. [Comissao Nacional de Eenrgia Nuclear, Rio de Janeiro (Brazil); Xavier, A. [PEN/Coppe -UFRJ, Rio de Janeiro (Brazil)

    2006-07-01

    Full text of publication follows: Nowadays, in a great number of situations constructions are raised around radiotherapy facilities. In cases where the constructions would not be in the primary x-ray beam, 'skyshine' radiation is normally accounted for. The skyshine method is commonly used to to calculate the dose contribution from scattered radiation in such circumstances, when the roof shielding is projected considering there will be no occupancy upstairs. In these cases, there will be no need to have the usual 1,5-2,0 m thick ceiling, and the construction costs can be considerably reduced. The existing expression to compute these doses do not accomplish to explain mathematically the existence of a shadow area just around the outer room walls, and its growth, as we get away from these walls. In this paper we propose a new method to compute photon skyshine doses, using geometrical considerations to find the maximum dose point. An empirical equation is derived, and its validity is tested using M.C.N.P. 5 Monte Carlo calculation to simulate radiotherapy rooms configurations. (authors)

  15. Analysis of the propagation of neutrons and gamma-rays from the fast neutron source reactor YAYOI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Shigeo, E-mail: neutron@keyaki.cc.u-tokai.ac.jp [Department of Energy Science and Engineering, School of Engineering, Tokai University, Hiratsuka, Kanagawa 259-1292 (Japan); Murata, Isao [Division of Electrical, Electronic and Information Engineering, Osaka University, Suita, Osaka 565-0871 (Japan); Nakagawa, Tsutomu; Saito, Isao [Nuclear Professional School, School of Engineering, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki 319-1188 (Japan)

    2011-10-01

    The skyshine effect is crucial for designing appropriate shielding. To investigate the skyshine effect, the propagation of neutrons was measured and analyzed at the fast neutron source reactor YAYOI. Pulse height spectra and dose distributions of neutron and secondary gamma-ray were measured outside YAYOI, and analyzed with MCNP-5 and JENDL-3.3. Comparison with the experimental results showed good agreement. Also, a semi-empirical formula was successfully derived to describe the dose distribution. The formulae can be used to predict the skyshine effect at YAYOI, and will be useful for estimating the skyshine effect and designing the shield structure for fusion facilities.

  16. Modification of SKYSHINE-III to include cask array shadowing

    Energy Technology Data Exchange (ETDEWEB)

    Hertel, N.E. [Georgia Institute of Technology, Atlanta, GA (United States); Pfeifer, H.J. [NAC International, Norcross, GA (United States); Napolitano, D.G. [NISYS Corporation, Duluth, GA (United States)

    2000-03-01

    The NAC International version of SKYSHINE-III has been expanded to represent the radiation emissions from ISFSI (Interim Spent Fuel Storage Installations) dry storage casks using surface source descriptions. In addition, this modification includes a shadow shielding algorithm of the casks in the array. The resultant code is a flexible design tool which can be used to rapidly assess the impact of various cask loadings and arrangements. An example of its use in calculating dose rates for a 10x8 cask array is presented. (author)

  17. Using MCNP code for neutron and photon skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)

  18. Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

    Science.gov (United States)

    Heuel-Fabianek, Burkhard; Hille, Ralf

    2005-01-01

    During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

  19. Impact of image noise on gamma index calculation

    International Nuclear Information System (INIS)

    Chen, M; Mo, X; Parnell, D; Olivera, G; Galmarini, D; Lu, W

    2014-01-01

    Purpose: The Gamma Index defines an asymmetric metric between the evaluated image and the reference image. It provides a quantitative comparison that can be used to indicate sample-wised pass/fail on the agreement of the two images. The Gamma passing/failing rate has become an important clinical evaluation tool. However, the presence of noise in the evaluated and/or reference images may change the Gamma Index, hence the passing/failing rate, and further, clinical decisions. In this work, we systematically studied the impact of the image noise on the Gamma Index calculation. Methods: We used both analytic formulation and numerical calculations in our study. The numerical calculations included simulations and clinical images. Three different noise scenarios were studied in simulations: noise in reference images only, in evaluated images only, and in both. Both white and spatially correlated noises of various magnitudes were simulated. For clinical images of various noise levels, the Gamma Index of measurement against calculation, calculation against measurement, and measurement against measurement, were evaluated. Results: Numerical calculations for both the simulation and clinical data agreed with the analytic formulations, and the clinical data agreed with the simulations. For the Gamma Index of measurement against calculation, its distribution has an increased mean and an increased standard deviation as the noise increases. On the contrary, for the Gamma index of calculation against measurement, its distribution has a decreased mean and stabilized standard deviation as the noise increases. White noise has greater impact on the Gamma Index than spatially correlated noise. Conclusions: The noise has significant impact on the Gamma Index calculation and the impact is asymmetric. The Gamma Index should be reported along with the noise levels in both reference and evaluated images. Reporting of the Gamma Index with switched roles of the images as reference and

  20. Impact of Image Noise on Gamma Index Calculation

    Science.gov (United States)

    Chen, M.; Mo, X.; Parnell, D.; Olivera, G.; Galmarini, D.; Lu, W.

    2014-03-01

    Purpose: The Gamma Index defines an asymmetric metric between the evaluated image and the reference image. It provides a quantitative comparison that can be used to indicate sample-wised pass/fail on the agreement of the two images. The Gamma passing/failing rate has become an important clinical evaluation tool. However, the presence of noise in the evaluated and/or reference images may change the Gamma Index, hence the passing/failing rate, and further, clinical decisions. In this work, we systematically studied the impact of the image noise on the Gamma Index calculation. Methods: We used both analytic formulation and numerical calculations in our study. The numerical calculations included simulations and clinical images. Three different noise scenarios were studied in simulations: noise in reference images only, in evaluated images only, and in both. Both white and spatially correlated noises of various magnitudes were simulated. For clinical images of various noise levels, the Gamma Index of measurement against calculation, calculation against measurement, and measurement against measurement, were evaluated. Results: Numerical calculations for both the simulation and clinical data agreed with the analytic formulations, and the clinical data agreed with the simulations. For the Gamma Index of measurement against calculation, its distribution has an increased mean and an increased standard deviation as the noise increases. On the contrary, for the Gamma index of calculation against measurement, its distribution has a decreased mean and stabilized standard deviation as the noise increases. White noise has greater impact on the Gamma Index than spatially correlated noise. Conclusions: The noise has significant impact on the Gamma Index calculation and the impact is asymmetric. The Gamma Index should be reported along with the noise levels in both reference and evaluated images. Reporting of the Gamma Index with switched roles of the images as reference and

  1. Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

    2004-07-01

    MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

  2. Development of Monte Carlo decay gamma-ray transport calculation system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

    2001-06-01

    In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

  3. Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

    International Nuclear Information System (INIS)

    Diop, C.M'B.

    1984-01-01

    This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

  4. LSHINSE, Air Scattering Neutron and Gamma Dose rates for Complex Shielding Geometry

    International Nuclear Information System (INIS)

    Baran, A.; Gruen, M.; Leicht, R.

    1991-01-01

    1 - Description of program or function: The program LSHINSE is used to calculate the flux and the dose rate caused by gamma radiation emanating from a point source and being scattered in surrounding air. The program considers all forms of single scattering. Multiple scattering is taken into account in an approximate way by use of buildup factors. 2 - Method of solution: The program LSHINSE solves the equations for skyshine by use of Simpson integration. The integration limits are chosen such that the partial shielding is approximated by rectangular walls around the source. In addition, the attenuation of the primary radiation by a room ceiling can be calculated for several materials. By giving the height of the ceiling, the scattering in the air of the room can be calculated. By specifying energy groups the spectrum of the scattered radiation can be obtained. Valid energy range is 0.1 - 0.2 MeV, where the lower limit is due to uncertainties in the buildup factors. 3 - Restrictions on the complexity of the problem: The program is restricted to rectangular shielding problems involving gamma radiation in the range of 0.1 to 2.0 MeV

  5. Secondary gamma-ray data for shielding calculation

    International Nuclear Information System (INIS)

    Miyasaka, Sunichi

    1979-01-01

    In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

  6. Analysis of neutron propagation from the skyshine port of a fusion neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Wakisaka, M. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan); Kaneko, J. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan)]. E-mail: kin@qe.eng.hokudai.ac.jp; Fujita, F. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan); Ochiai, K. [Japan Atomic Energy Institute, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Nishitani, T. [Japan Atomic Energy Institute, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Yoshida, S. [Tokai University, 1117 Kitakaname, Hirastuka, Kanagawa-ken 259-1292 (Japan); Sawamura, T. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan)

    2005-12-01

    The process of neutron leaking from a 14MeV neutron source facility was analyzed by calculations and experiments. The experiments were performed at the Fusion Neutron Source (FNS) facility of the Japan Atomic Energy Institute, Tokai-mura, Japan, which has a port on the roof for skyshine experiments, and a {sup 3}He counter surrounded with a polyethylene moderator of different thicknesses was used to estimate the energy spectra and dose distributions. The {sup 3}He counter with a 3-cm-thick moderator was also used for dose measurements, and the doses evaluated by the counter counts and the calculated count-to-dose conversion factor agreed with the calculations to within {approx}30%. The dose distribution was found to fit a simple analytical expression, D(r)=Q{sub D}exp(-r/{lambda}{sub D})r and the parameters Q{sub D} and {lambda}{sub D} are discussed.

  7. Observation of Neutron Skyshine from an Accelerator Based Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Franklyn, C. B. [Radiation Science Department, Necsa, PO Box 582, Pretoria 0001 (South Africa)

    2011-12-13

    A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

  8. Calculation method for gamma dose rates from Gaussian puffs

    Energy Technology Data Exchange (ETDEWEB)

    Thykier-Nielsen, S; Deme, S; Lang, E

    1995-06-01

    The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E{sub {gamma}}, {sigma}{sub y}, the asymmetry factor - {sigma}{sub y}/{sigma}{sub z}, the height of puff center - H and the distance from puff center R{sub xy}. To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs.

  9. COSANI-2, Gamma Doses from SABINE Calculation, Activity from ANISN Flux Calculation

    International Nuclear Information System (INIS)

    Dupont, C.

    1975-01-01

    1 - Nature of physical problem solved: Retrieval of SABINE and/or ANISN results. Calculates in case of SABINE results the individual contributions of capture gamma rays in each region to the total gamma dose and to the total gamma heating may calculate in case of ANISN new activity rates starting from ANISN flux saved on tape and activity cross sections taken on an ANISN binary library tape. The program can draw on a BENSON plotter any of the following quantities: - group flux; - activity rates; - dose rates; - neutron spectra for SABINE; - neutron or gamma direct or adjoint spectra for ANISN; - gamma heating and dose rate for SABINE including individual contributions from each region. Several ANISN and/or SABINE cases can be drawn on the same graph for comparison purposes. 2 - Restrictions on the complexity of the problem: Maximum number of: - tapes containing ANISN and/or SABINE results: 5; - curves per graph: 3; - regions: 40; - points per curve: 500; - energy groups: 200

  10. Dose contributions due to radiation scattered by air (skyshine) in the case of X-ray machines

    Energy Technology Data Exchange (ETDEWEB)

    Sahre, P.; Kaden, M. [Verein fuer Kernverfahrenstechnik und Analytik Rossendorf e.V. (VKTA), Dresden (Germany). Nuclear Engineering and Analytics Rossendorf; Schoenmuth, T.; Naumann, B. [Hochschule fuer Technik, Wirtschaft und Sozialwesen Zittau/Goerlitz, Zittau (Germany); Pawelke, J. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Technische Univ. Dresden (Germany); OncoRay, Dresden (Germany); Reichelt, U. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany)

    2012-06-15

    Radiation transport simulations had to be done in preparation of operation of the X-ray tube ISOVOLT 320 kV/13mA in a special laboratory. At first simulation was done without shielding the roof of the laboratory, showing a dose rate maximum of more than 100 mSv/h. This dose rate results in a skyshine dose rate of at most 2 {mu}Sv/h in the surrounding of the building without shielding the roof. For similar geometries the skyshine is negligible for dose rates at the unshielded roof of less than 3 mSv/h (exclusion area). (orig.)

  11. Calculation method for gamma dose rates from Gaussian puffs

    International Nuclear Information System (INIS)

    Thykier-Nielsen, S.; Deme, S.; Lang, E.

    1995-06-01

    The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E γ , σ y , the asymmetry factor - σ y /σ z , the height of puff center - H and the distance from puff center R xy . To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs

  12. Bulk density calculations from prompt gamma ray yield

    International Nuclear Information System (INIS)

    Naqvi, A.A.; Nagadi, M.M.; Al-Amoudi, O.S.B.; Maslehuddin, M.

    2006-01-01

    Full text: The gamma ray yield from a Prompt Gamma ray Neutron Activation Analysis (PGNAA) setup is a linear function of element concentration and neutron flux in a the sample with constant bulk density. If the sample bulk density varies as well, then the element concentration and the neutron flux has a nonlinear correlation with the gamma ray yield [1]. The measurement of gamma ray yield non-linearity from samples and a standard can be used to estimate the bulk density of the samples. In this study the prompt gamma ray yield from Blast Furnace Slag, Fly Ash, Silica Fumes and Superpozz cements samples have been measured as a function of their calcium and silicon concentration using KFUPM accelerator-based PGNAA setup [2]. Due to different bulk densities of the blended cement samples, the measured gamma ray yields have nonlinear correlation with calcium and silicon concentration of the samples. The non-linearity in the yield was observed to increase with gamma rays energy and element concentration. The bulk densities of the cement samples were calculated from ratio of gamma ray yield from blended cement and that from a Portland cement standard. The calculated bulk densities have good agreement with the published data. The result of this study will be presented

  13. Contribution to gamma ray transport calculation in heterogeneous media

    International Nuclear Information System (INIS)

    Bourdet, L.

    1985-04-01

    This thesis presents the development of gamma transport calculation codes in three dimension heterogeneous geometries. These codes allow us to define the protection against gamma-rays or verify their efficiency. The laws that govern the interactions of gamma-rays with matters are briefly revised. A library with the all necessary constants for these codes is created. TRIPOLI-2, a code that treats in exact way the neutron transport in matters using Monte-Carlo method, has been adapted to deal with the transport of gamma-rays in matters as well. TRINISHI, a code which considers only one collision, has been realized to treat heterogeneous geometries containing voids. Elaborating a formula that calculates the albedo for gamma-ray reflection (the code ALBANE) allows us to solve the problem of gamma-ray reflection on plane surfaces. NARCISSE-2 deals with gamma-rays that suffer only one reflection on the inner walls of any closed volume (rooms, halls...) [fr

  14. Applications to shielding design and others of monte carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Daiichiro [Mitsui Engineering and Shipbuiding Co., Ltd., Tokyo (Japan)

    2001-01-01

    One-dimensional or two-dimensional Sn computer code (ANISN, DOT3.5, etc.) and a point attenuation kernel integral code (QAD, etc.) have been used widely for shielding design. Application examples of monte carlo method which could follow precisely the three-dimensional configuration of shielding structure are shown as follow: (1) CASTER cask has a complex structure which consists of a large number of fuel baskets (stainless steel), neutron moderators (polyethylene rods), the body (cast iron), and cooling fin. The R-{theta} model of Sn code DOT3.5 cannot follow closely the complex form of polyethylene rods and fuel baskets. A monte carlo code MORSE is used to ascertain the calculation results of DOT3.5. The discrepancy between the calculation results of DOT3.5 and MORSE was in 10% for dose rate at distance of 1 m from the cask surface. (2) The dose rates of an iron cell at 10 cm above the floor are calculated by the code QAD and the MORSE. The reflected components of gamma ray caused by the auxiliary floor shield (lead) are analyzed by the MORSE. (3) A monte carlo code MCNP4A is used for skyshine evaluation of spent fuel carrier ship 'ROKUEIMARU'. The direct and skyshine components of gamma ray and neutron flux are estimated at each center of engine room and wheel house. The skyshine dose rate of neutron flux is 5-15 times larger than the gamma ray. (M. Suetake)

  15. Neutron skyshine measurement at a K1200 superconducting heavy ion cyclotron using bubble dosimeters

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, B. [Safety Div., Australian Nuclear Science and Technology Organisation, Menai (Australia); Ronningen, R.M. [Michigan State Univ., National Superconducting Cyclotron Lab., East Lansing, MI (United States); Rossi, P. [Michigan State Univ., Office of Radiation, Chemical and Biological Safety, East Lansing, MI (United States)

    1999-07-01

    Understanding the characteristics of the neutron skyshine radiation is necessary for an accurate assessment of the environmental dose in the vicinity of the containment of a high-energy particle accelerator. At the National Superconducting Cyclotron Laboratory (NSCL), neutron skyshine was measured, using beams of 140 MeV/nucleon {sup 4}He and 80 MeV/nucleon {sup 22}Ne ions from the K1200 superconducting cyclotron. After passing through a radioactive-beam production target, the ion beam stopped in a solid aluminium stopping bar inside of a dipole magnet, resulting in the production of high energy fragmentation as well as evaporation neutrons in the NSCL Analysis Hall. The neutron dose equivalent and energy spectrum at the 1.37 m thick concrete roof of the Analysis Hall, directly above the aluminium target bar (reference point), were estimated, using a spherical 'rem-counter' and a set of seven Bonner-spheres, respectively. The skyshine dose, from neutrons transmitted through 21.5-cm local iron 'shielding' of the dipole magnet and the concrete roof, were evaluated using superheated bubble dosimeters at 50 m, 75 m, 100 m and 115 m from the reference point. The neutron doses beyond the extremity of the NSCL facility were extrapolated from the results of this investigation and were used to predict the exposure to members of the public by considering the operation schedule of the K1200 cyclotron. (authors)

  16. Skyshine method for photons:comparison between theoretical approach and numerical simulation by Monte Carlo method; Skyshine para fotons: comparacao entre abordagem teorica e simulacao numerica pelo metodo de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Falcao, R.C.; Facure, A.; Santini, E.S. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)]. E-mail: ross@cnen.gov.br; afsoares@cnen.gov.br; esantini@cnen.gov.br; Silva, A.X. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: ademir@com.ufrj.br

    2005-07-01

    The skyshine method is commonly used to compute the dose around radiotherapy facilities, when the roof shielding is projected considering that there will be no occupancy upstairs. In these cases, there will be no need to have the usual 1,5-2,0 m thick ceiling, and the construction costs can be considerably reduced. The semi-empirical expression commonly used to compute these doses show a poor agreement with the experimental dose measurements found in the literature. In this paper the MCNP code was used to simulate the transport of photons in some radiotherapy rooms, with shielding projects approved by the Brazilian Nuclear Energy Commission (CNEN), and whose roof shielding were designed according to the above-mentioned method. These simulations are then compared with the calculations presented in the shielding projects and a clear discrepancy is observed between both results. (author)

  17. Neutron Skyshine in shielding projects of radiotherapy: comparison between theoretical approach and simulation by Monte Carlo method; 'Skyshine' de neutrons em projetos de blindagens de radioterapia: comparacao entre abordagem teorica e simulacao por metodo de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Falcao, R.C.; Facure, A. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Santini, E.S. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Silva, A.X. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

    2005-07-01

    In this work, the MCNP code is used to simulate the transport of neutrons in a room of radiotherapy, whose shieldings are designed according to the method of skyshine (scattering in the atmosphere). The simulations are compared with the results obtained from empirically established expressions, which are normally used for designing the ceilings of the rooms facilities, ensuring that dose rates (neutrons + photons) around them do not exceed the maximum limits allowed by the standards of the CNEN. Good agreement is observed between the doses calculated according to these expressions and those obtained through simulation by Monte Carlo in the case of rooms without ceiling, and an overestimate of the calculations by a factor 2 or 3 in relation to the simulations, in the case of rooms with ceiling.

  18. Calculation method for gamma-dose rates from spherical puffs

    International Nuclear Information System (INIS)

    Thykier-Nielsen, S.; Deme, S.; Lang, E.

    1993-05-01

    The Lagrangian puff-models are widely used for calculation of the dispersion of atmospheric releases. Basic output from such models are concentrations of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on semi-infinite cloud model. This method is however only applicable for points far away from the release point. The exact calculation of the cloud dose using the volume integral requires significant computer time. The volume integral for the gamma dose could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor due to the fact that the same correction factors are used for all isotopes. The authors describe a more elaborate correction method. This method uses precalculated values of the gamma-dose rate as a function of the puff dispersion parameter (δ p ) and the distance from the puff centre for four energy groups. The release of energy for each radionuclide in each energy group has been calculated and tabulated. Based on these tables and a suitable interpolation procedure the calculation of gamma doses takes very short time and is almost independent of the number of radionuclides. (au) (7 tabs., 7 ills., 12 refs.)

  19. ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

  20. Calculation of gamma-ray flux density above the Venus and Earth surfaces

    International Nuclear Information System (INIS)

    Surkov, Yu.A.; Manvelyan, O.S.

    1987-01-01

    Calculational results of dependence of flux density of nonscattered gamma-quanta on the height above the Venus and Earth planet surfaces are presented in the paper. Areas, where a certain part of gamma quanta is accumulated, are calaculted for each height. Spectra of scattered gamma quanta and their integral fluxes at different heights above the Venera planet surface are calculated. Effect of the atmosphere on gamma radiation recorded is considered. The results obtained allow to estimate optimal conditions for measuring gamma-fields above the Venus and Earth planet surfaces, to determine the area of the planet surface investigated. They are also necessary to determine the elementary composition of the rock according to the characteristic gamma radiation spectrum recorded

  1. The Monte Carlo calculation of gamma family

    International Nuclear Information System (INIS)

    Shibata, Makio

    1980-01-01

    The method of the Monte Carlo calculation for gamma family was investigated. The effects of the variation of values or terms of parameters on observed quantities were studied. The terms taken for the standard calculation are the scaling law for the model, simple proton spectrum for primary cosmic ray, a constant cross section of interaction, zero probability of neutral pion production, and the bending of the curve of primary energy spectrum. This is called S model. Calculations were made by changing one of above mentioned parameters. The chamber size, the mixing of gamma and hadrons, and the family size were fitted to the practical ECC data. When the model was changed from the scaling law to the CKP model, the energy spectrum of the family was able to be expressed by the CKP model better than the scaling law. The scaling law was better in the symmetry around the family center. It was denied that primary cosmic ray mostly consists of heavy particles. The increase of the interaction cross section was necessary in view of the frequency of the families. (Kato, T.)

  2. Calculating gamma dose factors for hot particle exposures

    International Nuclear Information System (INIS)

    Murphy, P.

    1990-01-01

    For hot particle exposures to the skin, the beta component of radiation delivers the majority of the dose. However, in order to fully demonstrate regulatory compliance, licenses must ordinarily provide reasonable bases for assuming that both the gamma component of the skin dose and the whole body doses are negligible. While beta dose factors are commonly available in the literature, gamma dose factors are not. This paper describes in detail a method by which gamma skin dose factors may be calculated using the Specific Gamma-ray Constant, even if the particle is not located directly on the skin. Two common hot particle exposure geometries are considered: first, a single square centimeter of skin lying at density thickness of 7 mg/cm 2 and then at 1000 mg/cm 2 . A table provides example gamma dose factors for a number of isotopes encountered at power reactors

  3. Methodology comparison for gamma-heating calculations in material-testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)

    2015-07-01

    The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear

  4. Calculation and design for SSRF's bulk shield

    Energy Technology Data Exchange (ETDEWEB)

    Fang, K.M. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)]. E-mail: fangkm@sinap.ac.cn; Xu, X.J. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China); Cai, J.H. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)

    2006-12-15

    Shielding design objectives for the SSRF are chosen, assumptions for beam loss rates are given, the methods used on the APS by Moe are summarized and introduced to make calculation and design on bulk shield, the factor of skyshine is also considered, design thicknesses for SSRF's bulk shield are presented.

  5. A formalism for independent checking of Gamma Knife dose calculations

    International Nuclear Information System (INIS)

    Tsai Jensan; Engler, Mark J.; Rivard, Mark J.; Mahajan, Anita; Borden, Jonathan A.; Zheng Zhen

    2001-01-01

    For stereotactic radiosurgery using the Leksell Gamma Knife system, it is important to perform a pre-treatment verification of the maximum dose calculated with the Leksell GammaPlan[reg] (D LGP ) stereotactic radiosurgery system. This verification can be incorporated as part of a routine quality assurance (QA) procedure to minimize the chance of a hazardous overdose. To implement this procedure, a formalism has been developed to calculate the dose D CAL (X,Y,Z,d av ,t) using the following parameters: average target depth (d av ), coordinates (X,Y,Z) of the maximum dose location or any other dose point(s) to be verified, 3-dimensional (3-dim) beam profiles or off-center-ratios (OCR) of the four helmets, helmet size i, output factor O i , plug factor P i , each shot j coordinates (x,y,z) i,j , and shot treatment time (t i,j ). The average depth of the target d av was obtained either from MRI/CT images or ruler measurements of the Gamma Knife Bubble Head Frame. D CAL and D LGP were then compared to evaluate the accuracy of this independent calculation. The proposed calculation for an independent check of D LGP has been demonstrated to be accurate and reliable, and thus serves as a QA tool for Gamma Knife stereotactic radiosurgery

  6. EBT-P gamma-ray-shielding analysis

    International Nuclear Information System (INIS)

    Gohar, Y.

    1983-01-01

    First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

  7. A gamma heating calculation methodology for research reactor application

    International Nuclear Information System (INIS)

    Lee, Y.K.; David, J.C.; Carcreff, H.

    2001-01-01

    Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

  8. GAMMA-CLOUD: a computer code for calculating gamma-exposure due to a radioactive cloud released from a point source

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, O [Chugoku Electric Power Co. Inc., Hiroshima (Japan); Sawaguchi, Y; Kaneko, M

    1979-03-01

    A computer code, designated GAMMA-CLOUD, has been developed by specialists of electric power companies to meet requests from the companies to have a unified means of calculating annual external doses from routine releases of radioactive gaseous effluents from nuclear power plants, based on the Japan Atomic Energy Commission's guides for environmental dose evaluation. GAMMA-CLOUD is written in FORTRAN language and its required capacity is less than 100 kilobytes. The average ..gamma..-exposure at an observation point can be calculated within a few minutes with comparable precision to other existing codes.

  9. Efficient gamma index calculation using fast Euclidean distance transform

    Energy Technology Data Exchange (ETDEWEB)

    Chen Mingli; Lu Weiguo; Chen Quan; Ruchala, Kenneth; Olivera, Gustavo [TomoTherapy Inc., 1240 Deming Way, Madison, WI 53717 (United States)], E-mail: wlu@tomotherapy.com

    2009-04-07

    The gamma index is a tool for dose distribution comparison. It combines both dose difference (DD) and distance to agreement (DTA) into a single quantity. Though it is an effective measure, making up for the inadequacy of DD or DTA alone, its calculation can be very time-consuming. For a k-D space with N quantization levels in each dimension, the complexity of the exhaustive search is O(N{sup 2k}). In this work, we proposed an efficient method that reduces the complexity from O(N{sup 2k}) to O(N{sup k}M), where M is the number of discretized dose values and is comparable to N. More precisely, by embedding the reference dose distribution in a (k+1)-D spatial-dose space, we can use fast Euclidean distance transform with linear complexity to obtain a table of gamma indices evaluated over a range of the (k+1)-D spatial-dose space. Then, to obtain gamma indices for the test dose distribution, it requires only table lookup with complexity O(N{sup k}). Such a table can also be used for other test dose distributions as long as the reference dose distribution is the same. Simulations demonstrated the efficiency of our proposed method. The speedup for 3D gamma index calculation is expected to be on the order of tens of thousands (from O(N{sup 6}) to O(N{sup 3}M)) if N is a few hundreds, which makes clinical usage of the 3D gamma index feasible. A byproduct of the gamma index table is that the gradient of the gamma index with respect to either the spatial or dose dimension can be easily derived. The gradient can be used to identify the main causes of the discrepancy from the reference distribution at any dose point in the test distribution or incorporated in treatment planning and machine parameter optimization.

  10. Calculation of the secondary gamma radiation by the Monte Carlo method at displaced sampling from distributed sources

    International Nuclear Information System (INIS)

    Petrov, Eh.E.; Fadeev, I.A.

    1979-01-01

    A possibility to use displaced sampling from a bulk gamma source in calculating the secondary gamma fields by the Monte Carlo method is discussed. The algorithm proposed is based on the concept of conjugate functions alongside the dispersion minimization technique. For the sake of simplicity a plane source is considered. The algorithm has been put into practice on the M-220 computer. The differential gamma current and flux spectra in 21cm-thick lead have been calculated. The source of secondary gamma-quanta was assumed to be a distributed, constant and isotropic one emitting 4 MeV gamma quanta with the rate of 10 9 quanta/cm 3 xs. The calculations have demonstrated that the last 7 cm of lead are responsible for the whole gamma spectral pattern. The spectra practically coincide with the ones calculated by the ROZ computer code. Thus the algorithm proposed can be offectively used in the calculations of secondary gamma radiation transport and reduces the computation time by 2-4 times

  11. Calculation Analysis of Calibration Factors of Airborne Gamma-ray Spectrometer

    International Nuclear Information System (INIS)

    Zhao Jun; Zhu Jinhui; Xie Honggang; He Qinglin

    2009-01-01

    To determine the calibration factors of an airborne gamma-ray spectrometer measuring large area gamma-ray emitting source at deferent flying height, a series of Monte Carlo simulations were drawn. Response energy spectrums of NaI crystals in airplane caused by nature-decay-series calibration-pads, and calibration factors on different heights above Cs-137 plane source, were obtained. The calculated results agreed with the experimental data well. (authors)

  12. Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

    International Nuclear Information System (INIS)

    Lefvert, T.

    1975-11-01

    Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

  13. Measurement and Calculation of Gamma Radiation from HWZPR Reactor

    International Nuclear Information System (INIS)

    Jalali, Majid

    2006-01-01

    HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)

  14. Calculation of the flux density of gamma rays above the surface of Venus and the Earth

    International Nuclear Information System (INIS)

    Surkov, Yu.A.; Manvelyan, O.S.

    1987-01-01

    In this article the authors present the results of calculating the flux density of unscattered gamma rays as a function of height above the surfaces of Venus and the Earth. At each height they calculate the areas which will collect a certain fraction of the gamma rays. The authors calculate the spectra of scattered gamma rays, as well as their integrated fluxes at various heights above the surface of Venus. They consider how the atmosphere will affect the recording of gamma rays. Their results enable them to evaluate the optimal conditions for measuring the gamma-ray fields above the surfaces of Venus and the Earth and to determine the area of the planet which can be investigated in this way. These results are also necessary if they are to determine the elemental composition of the rock from the characteristic recorded spectrum of gamma radiation

  15. Simplified shielding calculation system for high-intensity proton accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Masumura, Tomomi; Nakashima, Hiroshi; Nakane, Yoshihiro; Sasamoto, Nobuo [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-06-01

    A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)

  16. Dose Distribution Calculation Using MCNPX Code in the Gamma-ray Irradiation Cell

    International Nuclear Information System (INIS)

    Kim, Yong Ho

    1991-02-01

    60 Co-gamma irradiators have long been used for foods sterilization, plant mutation and development of radio-protective agents, radio-sensitizers and other purposes. The Applied Radiological Science Research Institute of Cheju National University has a multipurpose gamma irradiation facility loaded with a MDS Nordin standard 60 Co source (C188), of which the initial activity was 400 TBq (10,800 Ci) on February 19, 2004. This panoramic gamma irradiator is designed to irradiate in all directions various samples such as plants, cultured cells and mice to administer given radiation doses. In order to give accurate doses to irradiation samples, appropriate methods of evaluating, both by calculation and measurement, the radiation doses delivered to the samples should be set up. Computational models have been developed to evaluate the radiation dose distributions inside the irradiation chamber and the radiation doses delivered to typical biolological samples which are frequently irradiated in the facility. The computational models are based on using the MCNPX code. The horizontal and vertical dose distributions has been calculated inside the irradiation chamber and compared the calculated results with measured data obtained with radiation dosimeters to verify the computational models. The radiation dosimeters employed are a Famer's type ion chamber and MOSFET dosimeters. Radiation doses were calculated by computational models, which were delivered to cultured cell samples contained in test tubes and to a mouse fixed in a irradiation cage, and compared the calculated results with the measured data. The computation models are also tested to see if they can accurately simulate the case where a thick lead shield is placed between the source and detector. Three tally options of the MCNPX code, F4, F5 and F6, are alternately used to see which option produces optimum results. The computation models are also used to calculate gamma ray energy spectra of a BGO scintillator at

  17. PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

    International Nuclear Information System (INIS)

    Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

    1990-01-01

    1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

  18. Calculation of gamma spectra for positron annihilation on molecules

    Energy Technology Data Exchange (ETDEWEB)

    Green, Dermot G; Gribakin, G F [Centre for Theoretical Atomic, Molecular and Optical Physics, Queen' s University Belfast, BT7 1NN (United Kingdom); Wang, F [Centre for Molecular Simulation, Sinburne University of Technology, Melbourne, Victoria 3122 (Australia); Surko, C M, E-mail: dgreen09@qub.ac.u [Physics Department, University of California, San Diego, La Jolla, California 92093-0319 (United States)

    2009-11-01

    Gamma spectra for positron annihilation on molecules are calculated based on molecular electron momentum densities and using an atomic adjustment factor that accounts for the positron. Results for H{sub 2} agree well with experiment. Analysis of methane and larger alkanes and their substitutes is underway.

  19. Shielding calculations for the Intense Neutron Source Facility. Final report

    International Nuclear Information System (INIS)

    Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

    1978-06-01

    Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

  20. Demonstration study on shielding safety analysis code (8)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  1. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  2. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  3. Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

    International Nuclear Information System (INIS)

    Kwon, S.G.; Lee, S.Y.; Yook, C.C.

    1981-01-01

    This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

  4. Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Nimal, J.C.

    1982-01-01

    This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements

  5. Calculation of Dose Gamma Ray Build up Factor in Some ...

    African Journals Online (AJOL)

    The gamma ray buildup factor was calculated by analyzing the narrow- beam and broad-beam geometry equations using Taylor's formula for isotropic sources and homogeneous materials. The buildup factor was programmed using MATLAB software to operate with any radiation energy (E), atomic number (Z) and the ...

  6. Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique

    International Nuclear Information System (INIS)

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

    2009-01-01

    In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.

  7. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  8. Isodose distributions and dose uniformity in the Portuguese gamma irradiation facility calculated using the MCNP code

    CERN Document Server

    Oliveira, C

    2001-01-01

    A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.

  9. Use of integral experiments to improve neutron propagation and gamma heating calculations

    International Nuclear Information System (INIS)

    Oceraies, Y.; Caumette, P.; Devillers, C.; Bussac, J.

    1979-01-01

    1) The studies to define and improve the accuracies of neutron propagation and gamma heating calculations from integral experiments are encompassed in the field of the fast reactor physics program at CEA. 2) A systematic analysis of neutron propagation in Fe-Na clean media, with variable volumic composition between 0 and 100% in sodium, has been performed on the HARMONIE source reactor. Gamma heating traverses in the core, the blankets and several control rods, have been measured in the R Z core program at MASURCA. The experimental techniques, the accuracies and the results obtained are given. The approximations of the calculational methods used to analyse these experiments and to predict the corresponding design parameters are also described. 3) Particular emphasis is given to the methods planned to improve fundamental data used in neutron propagation calculations, using the discrepancies observed between measured and calculated results in clean integral experiments. One of these approaches, similar to the techniques used in core physics, relies upon sensitivity studies and eventually on adjustment techniques applied to neutron propagation. (author)

  10. Implementing displacement damage calculations for electrons and gamma rays in the Particle and Heavy-Ion Transport code System

    Science.gov (United States)

    Iwamoto, Yosuke

    2018-03-01

    In this study, the Monte Carlo displacement damage calculation method in the Particle and Heavy-Ion Transport code System (PHITS) was improved to calculate displacements per atom (DPA) values due to irradiation by electrons (or positrons) and gamma rays. For the damage due to electrons and gamma rays, PHITS simulates electromagnetic cascades using the Electron Gamma Shower version 5 (EGS5) algorithm and calculates DPA values using the recoil energies and the McKinley-Feshbach cross section. A comparison of DPA values calculated by PHITS and the Monte Carlo assisted Classical Method (MCCM) reveals that they were in good agreement for gamma-ray irradiations of silicon and iron at energies that were less than 10 MeV. Above 10 MeV, PHITS can calculate DPA values not only for electrons but also for charged particles produced by photonuclear reactions. In DPA depth distributions under electron and gamma-ray irradiations, build-up effects can be observed near the target's surface. For irradiation of 90-cm-thick carbon by protons with energies of more than 30 GeV, the ratio of the secondary electron DPA values to the total DPA values is more than 10% and increases with an increase in incident energy. In summary, PHITS can calculate DPA values for all particles and materials over a wide energy range between 1 keV and 1 TeV for electrons, gamma rays, and charged particles and between 10-5 eV and 1 TeV for neutrons.

  11. On calculation of detection efficiency of gamma spectrometers with germanium detection

    International Nuclear Information System (INIS)

    Sima, O.

    2001-01-01

    High resolution gamma spectrometer represents a powerful analysis technique of use in various fields from basic research to the study of environmental radioactivity, from medical investigations to geological surveys. Direct experimental calibration cannot cover the large range of measurement configurations of interest. Actually, it can be appropriately applied in an only limited number of cases, as for instance, in case of point-like sources or liquid phase volume sources. To assist the treatment of experimental calibration of germanium detectors, in the frame of Atomic and Nuclear Physics Chair of Department of Physics, a number of calculation methods were developed. These methods are generally based on Monte Carlo simulation but simplified and fast analytical methods were also worked out. Initially, these studies were dedicated to application in the field of environmental activity and radiation protection, but later on these were extended also to other fields as, for instance, the neutron activation or radionuclide metrology. First, the effects of matrices were calculated for the case of volume sources. Applying the matrix corrections allows obtaining the source calibration curves on the basis of experimental calibration data obtained with liquid sources, in the same geometry. An algorithm based on Monte Carlo calculation and using techniques of correlated selection was obtained. This algorithm can be implemented in the gamma analysis programs giving for the first time the possibility of correct evaluation of matrix effects even during the analysis of gamma spectra. We used a set of additive relations applicable in case of volume sources with negligible self-absorption and obtained a number of linear relations useful in calibrating the large volume sources in presence of self-absorption, based on small volume standard sources. Also, we proposed analytical relations useful in the case of measurements of large volume samples, in case of Marinelli geometry. To

  12. Study on dose assessment in surrounding environment of the Tono Mine associated with closure activity

    International Nuclear Information System (INIS)

    Sasao, Eiji

    2012-07-01

    Dose assessment associated with closure activity of the Tono Mine has been performed. In this assessment, exposure dose has been calculated on groundwater and surface water migration of radionuclide from 1) waste rock in the waste rock dump facility, 2) mining waste in the mining waste facility, and 3) uranium ore and waste rock backfilled in the shafts and galleries. Direct and skyshine gamma rays and exposure of exhalated radon from the waste rock dump has also been evaluated. An evaluation tool developed for safety assessment for sub-surface disposal of radioactive waste is utilized for this assessment. Localities for dose evaluation are selected at the Higashihoragawa and Hiyoshigawa based on the topography around the Tono Mine and groundwater flow simulation. Evaluation scenarios are classified into 'Scenario for intake of agricultural product' as the base scenario, and 'Scenario for intake of groundwater' as the alternative scenario. Parameters for dose assessment are set-up based on the existing data. But the range and uncertainty of parameters are taken into account in the 'alternative cases'. As the result of dose assessment, maximum exposure dose of the base scenario is 0.08mSv/year, and 0.09mSv/year including direct and skyshine gamma rays and exposure of exhalatedradon at the Higashihoragawa. Maximum exposure dose of the alternative scenario is 0.08mSv/year (0.09mSv/year including direct and skyshine gamma rays and exposure of exhalated radon). On the alternative cases, exposure doses are calculated as 0.05-0.14mSv/year in both of the base and alternative scenarios. At the Hiyoshigawa, maximum exposure dose is less than 0.001mSv/year (1x10 -6 mSv/year) for the base scenario, and 0.001mSv/year for the alternative scenario. On the alternative cases, maximum exposure doses are less than 0.001mSv/year for all cases of the base scenario and 0.0006-0.002mSv/year for the alternative scenario. (author)

  13. Neutrons leaked from a 45 MeV linac facility

    Energy Technology Data Exchange (ETDEWEB)

    Kitaichi, Masatoshi; Sawamura, Sadashi; Yamada, Takuma; Sawamura, Teruko; Kaneko, Junnichi H. [Hokkaido Univ., Sapporo (Japan); Nojiri, Itiro [Japan Nuclear Cycle Development Institute, Ibaraki (Japan)

    2002-07-01

    Dose evaluation for skyshine from nuclear facilities is an issue in environmental evaluations. Therefore, benchmark data for skyshine and well-investigated codes for skyshine would be useful in the rational evaluations of nuclear facilities. The purpose of this study is to obtain benchmark data of skyshine and to investigate the effect of source spectra and angular distribution on the skyshine process. In this study spatial and time distributions of neutrons leaked from the Hokkaido University 45 MeV electron linac facility were measured and compared with calculations. Neutrons were emitted from the ( ,n) reaction produced by bremsstrahlung radiation in a lead target irradiated with electrons from the linac. The skyshine process of neutrons transported through the facility building to the outside was investigated. The source spectrum of the skyshine process was evaluated using a cylindrical multi-moderator spectrometer and unfolding code, the SAND-II, and the results were compared. Measurements were carried out to a distance of 330 m from the facility. The measured spatial dose distribution was found not to coincide with the calculations. The discrepancy is discussed based on an analysis of the spatial and time distributions, and the energy spectrum which suggests that the source spectrum and the angular distribution assumed in the calculation was not sufficiently similar to simulate the experimental situation. The time distribution introduced in this study appears to be useful in discussions of the skyshine process and its sources.

  14. Comparison of a semi-empirical method with some model codes for gamma-ray spectrum calculation

    Energy Technology Data Exchange (ETDEWEB)

    Sheng, Fan; Zhixiang, Zhao [Chinese Nuclear Data Center, Beijing, BJ (China)

    1996-06-01

    Gamma-ray spectra calculated by a semi-empirical method are compared with those calculated by the model codes such as GNASH, TNG, UNF and NDCP-1. The results of the calculations are discussed. (2 tabs., 3 figs.).

  15. Shielding analysis of high level waste water storage facilities using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Yabuta, Naohiro [Mitsubishi Research Inst., Inc., Tokyo (Japan)

    2001-01-01

    The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)

  16. Measurement and calculation of characteristic prompt gamma ray spectra emitted during proton irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Polf, J C; Peterson, S; Beddar, S [M D Anderson Cancer Center, Univeristy of Texas, Houston, TX 77030 (United States); McCleskey, M; Roeder, B T; Spiridon, A; Trache, L [Cyclotron Institute, Texas A and M University, College Station, TX 77843 (United States)], E-mail: jcpolf@mdanderson.org

    2009-11-21

    In this paper, we present results of initial measurements and calculations of prompt gamma ray spectra (produced by proton-nucleus interactions) emitted from tissue equivalent phantoms during irradiations with proton beams. Measurements of prompt gamma ray spectra were made using a high-purity germanium detector shielded either with lead (passive shielding), or a Compton suppression system (active shielding). Calculations of the spectra were performed using a model of both the passive and active shielding experimental setups developed using the Geant4 Monte Carlo toolkit. From the measured spectra it was shown that it is possible to distinguish the characteristic emission lines from the major elemental constituent atoms (C, O, Ca) in the irradiated phantoms during delivery of proton doses similar to those delivered during patient treatment. Also, the Monte Carlo spectra were found to be in very good agreement with the measured spectra providing an initial validation of our model for use in further studies of prompt gamma ray emission during proton therapy. (note)

  17. Calculation of gamma ray exposure rates from uranium ore bodies

    International Nuclear Information System (INIS)

    Thomson, J.E.; Wilson, O.J.

    1980-02-01

    The planning of operations associated with uranium mines often requires that estimates be made of the exposure rates from various ore bodies. A straight-forward method of calculating the exposure rate from an arbitrarily shaped body is presented. Parameters for the calculation are evaluated under the assumption of secular equilibrium of uranium with its daughters and that the uranium is uniformly distributed throughout an average soil mixture. The spectral distribution of the emitted gamma rays and the effect of air attenuation are discussed. Worked examples are given of typical situations encountered in uranium mines

  18. Calorific energy deposited by gamma radiations in a test reactor. Calorimetric measurements and calculations

    International Nuclear Information System (INIS)

    Mecheri, K.-F.

    1977-01-01

    The purpose of this work was to determine the calorific energy deposited by gamma radiations in the experimental devices irradiated in the test reactors of the Grenoble Nuclear Study Centre. A theoretical study briefly recalls to mind the various sorts of nuclear reactions that occur in a reactor, from the special angle of their ability to deposit calorific energy in the materials. A special study with the help of a graphite calorimeter made it possible to show the possible effect of the various parameters intervening in this energy absorption: the nature of the materials, their geometry, the spectrum of the incident gamma rays and the fact that the variation of this spectrum is due to the position of the measuring point with respect to the reactor core or to the presence of structures around the measuring instrument. The results of the calculations made with the help of the Mercury IV and ANISN codes are compared with those of the determinations in order to ascertain that very are adapted to the forecasts of energy deposition in the various materials. The conclusion was reached that in order to calculate with accuracy the depositifs of gamma energy in the experimental devices, it is necessary either to introduce the build-up calculation for the low energy photons, in the Mercury IV calculation code or to associate the DOT code to the ANISN calculation code [fr

  19. Poster - 11: Radiation barrier thickness calculations for the GammaPod

    International Nuclear Information System (INIS)

    La Russa, Daniel; Vandervoort, Eric; Wilkins, David

    2016-01-01

    A consortium of radiotherapy centers in North America is in the process of evaluating a novel new 60 Co teletherapy device, called the GammaPod™ (Xcision Medical Systems, Columbia Maryland), designed specifically for breast SBRT. The GammaPod consists of 36 collimated 60 Co sources with a total activity of 4320 Ci. The sources are housed in a hemispherical source carrier that rotates during treatment to produce a cylindrically symmetric cone of primary beam spanning 16° – 54° degrees from the horizontal. This unique beam geometry presents challenges when designing or evaluating room shielding for the purposes of meeting regulatory requirements, and for ensuring the safety of staff and the public in surrounding areas. Conventional methods for calculating radiation barrier thicknesses have been adapted so that barrier transmission factors for the GammaPod can be determined from a few relevant distances and characteristics of the primary beam. Simple formalisms have been determined for estimating shielding requirements for primary radiation (with a rotating and non-rotating source carrier), patient-scattered radiation, and leakage radiation. When making worst case assumptions, it was found that conventional barrier thicknesses associated with linac treatment suites are sufficient for shielding all sources of radiation from the GammaPod.

  20. Poster - 11: Radiation barrier thickness calculations for the GammaPod

    Energy Technology Data Exchange (ETDEWEB)

    La Russa, Daniel; Vandervoort, Eric; Wilkins, David [Radiation Medicine Program, The Ottawa Hospital (Canada)

    2016-08-15

    A consortium of radiotherapy centers in North America is in the process of evaluating a novel new {sup 60}Co teletherapy device, called the GammaPod™ (Xcision Medical Systems, Columbia Maryland), designed specifically for breast SBRT. The GammaPod consists of 36 collimated {sup 60}Co sources with a total activity of 4320 Ci. The sources are housed in a hemispherical source carrier that rotates during treatment to produce a cylindrically symmetric cone of primary beam spanning 16° – 54° degrees from the horizontal. This unique beam geometry presents challenges when designing or evaluating room shielding for the purposes of meeting regulatory requirements, and for ensuring the safety of staff and the public in surrounding areas. Conventional methods for calculating radiation barrier thicknesses have been adapted so that barrier transmission factors for the GammaPod can be determined from a few relevant distances and characteristics of the primary beam. Simple formalisms have been determined for estimating shielding requirements for primary radiation (with a rotating and non-rotating source carrier), patient-scattered radiation, and leakage radiation. When making worst case assumptions, it was found that conventional barrier thicknesses associated with linac treatment suites are sufficient for shielding all sources of radiation from the GammaPod.

  1. Actuarial values calculated using the incomplete Gamma function

    Directory of Open Access Journals (Sweden)

    Giovanni Mingari Scarpello

    2013-03-01

    Full Text Available The complete expectation-of-life for a person and the actuarial present value of continuous life annuities are defined by integrals. In all of them at least one of the factors is a survival function value ratio. If de Moivre’s law of mortality is chosen, such integrals can easily be evaluated; but if the Makeham survival function is adopted, they are used to be calculated numerically. For the above actuarial figures, closed form integrations are hereafter provided by means of the incomplete Gamma function.

  2. Calculation of Buildup Factor for Gamma-ray Exposure in Two Layered Shields Made of Water and Lead

    International Nuclear Information System (INIS)

    Al-Saadi, A.H.

    2012-01-01

    The buildup factor for gamma ray exposure is most useful in calculations for biological protective shields.The buildup factors for gamma ray exposure were calculated in tow layered shields consist of water-lead and lead-water up to optical Thickness 20 mean free path (mfp) at gamma ray energies 1, 2 and 6MeV by using kalos's formula.The program has been designed to work at any atomic number of the attenuating medium, photon energy, slab thickness and and the arrangement of materials.The results obtained in this search leading to the buildup factor for gamma ray exposure at energies (1and2MeV) in lead-water were higher than the reverse case,while at energy 6 MeV the effect was opposite.The calculated data were parameterized by an empirical formula as a function of optical thickness of tow materials.The results obtained were in reasonable agreement with a previous work

  3. Calculation of point isotropic buildup factors of gamma rays for water and lead

    Directory of Open Access Journals (Sweden)

    A. S. H.

    2001-12-01

    Full Text Available   Exposure buildup factors for water and lead have been calculated by the Monte-Carlo method for an isotropic point source in an infinite homogeneous medium, using the latest cross secions available on the Internet. The types of interactions considered are ,photoelectric effect, incoherent (or bound-electron Compton. Scattering, coherent (or Rayleigh scattering and pair production. Fluorescence radiations have also been taken into acount for lead. For each material, calculations were made at 10 gamma ray energies in the 40 keV to 10 MeV range and up to penetration depths of 10 mean free paths at each energy point. The results presented in this paper can be considered as modified gamma ray exposure buildup factors and be used in radiation shielding designs.

  4. Improved calculation of displacements per atom cross section in solids by gamma and electron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Piñera, Ibrahin, E-mail: ipinera@ceaden.edu.cu [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Cruz, Carlos M.; Leyva, Antonio; Abreu, Yamiel; Cabal, Ana E. [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Espen, Piet Van; Remortel, Nick Van [University of Antwerp, CGB, Groenenborgerlaan 171, 2020 Antwerpen (Belgium)

    2014-11-15

    Highlights: • We present a calculation procedure for dpa cross section in solids under irradiation. • Improvement about 10–90% for the gamma irradiation induced dpa cross section. • Improvement about 5–50% for the electron irradiation induced dpa cross section. • More precise results (20–70%) for thin samples irradiated with electrons. - Abstract: Several authors had estimated the displacements per atom cross sections under different approximations and models, including most of the main gamma- and electron-material interaction processes. These previous works used numerical approximation formulas which are applicable for limited energy ranges. We proposed the Monte Carlo assisted Classical Method (MCCM), which relates the established theories about atom displacements to the electron and positron secondary fluence distributions calculated from the Monte Carlo simulation. In this study the MCCM procedure is adapted in order to estimate the displacements per atom cross sections for gamma and electron irradiation. The results obtained through this procedure are compared with previous theoretical calculations. An improvement in about 10–90% for the gamma irradiation induced dpa cross section is observed in our results on regard to the previous evaluations for the studied incident energies. On the other hand, the dpa cross section values produced by irradiation with electrons are improved by our calculations in about 5–50% when compared with the theoretical approximations. When thin samples are irradiated with electrons, more precise results are obtained through the MCCM (in about 20–70%) with respect to the previous studies.

  5. Demonstration study on shielding safety analysis code. 7

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    2000-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a {sup 3}He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

  6. Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Nojiri, Ichiro; Odajima, Akira; Sasaki, Toshihisa; Kurosawa, Naohiro

    1998-01-01

    In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)

  7. For a Better Estimation of Gamma Heating in Experimental Reactors and Devices: Stakes and Work Plan from Calculation Methods to Nuclear Data

    International Nuclear Information System (INIS)

    Lemaire, Matthieu; Vaglio-Gaudard, Claire; Lyoussi, Abdallah; Reynard-Carette, Christelle

    2013-06-01

    The Jules Horowitz Reactor (JHR) is an international Material-Testing Reactor currently under construction at CEA Cadarache. The determination of gamma heating levels in this future commercial reactor is of crucial importance as gamma heating affects both safety and performance parameters of the JHR. Required accuracy (5% at one standard deviation) makes it necessary to calibrate bias and uncertainty associated with JHR gamma-heating calculations. Main steps of bias determination for gamma-heating calculations include, firstly, the development of a calculation methodology with controlled use of physical approximations; secondly, the interpretation of gamma-heating measurements so as to determine bias supposed to be only due to nuclear data. (authors)

  8. A new approximating formula for calculating gamma-ray buildup factors in multilayer shields

    International Nuclear Information System (INIS)

    Assad, A.; Chiron, M.; Nimal, J.C.; Diop, C.M.; Ridoux, P.

    1999-01-01

    This study proposes a new approximating formula for calculating gamma-ray buildup factors in multilayer shields. The formula combines the buildup factors of single-layer shields with products and quotients. The feasibility of the formula for reproducing the buildup factors was tested by using point isotropic buildup factors calculated with the SN1D discrete ordinates code as reference data. The dose buildup factors of single-, double-, and multilayer shields composed of water, aluminum, iron, and lead were calculated for a spherical geometry in the energy range between 10 MeV and 40 keV and for total thicknesses of up to 30 mean free paths. The calculation of the buildup factors takes into account the bound electron effect of Compton scattering (incoherent scattering), the coherent scattering, the pair production, and the secondary sources of bremsstrahlung and fluorescence. The tests have shown that the approximating formula reproduces the reference data of double-layer shields very well for most cases. With the same parameters and with a new physical consideration that takes into account in a global way the degradation of the gamma-ray energy spectrum, the buildup factors of three- and five-layer shields were also very well reproduced

  9. Verification calculation of drum and pulley overhead travelling crane on gamma irradiators

    International Nuclear Information System (INIS)

    Syamsurrijal Ramdja; Ari Satmoko; Sutomo Budihardjo

    2010-01-01

    Having verified the calculation of dam drum pulleys found on cranes to facilitate the gamma irradiator. Drum is a device for rolling steel ropes while the pulley is a circular pieces called disks, which are made from metal or non-metal to transmit motion and force. Having verified calculation of forces acting style on drums, drum diameter and length and style of press that occurred on drums. Likewise, the pulley, pulley diameter verified calculations, measures of disc and shaft power pulleys. From the verification results will be obtained whether the data drums and pulley device is safe or not safe to use. (author)

  10. ParShield: A computer program for calculating attenuation parameters of the gamma rays and the fast neutrons

    International Nuclear Information System (INIS)

    Elmahroug, Y.; Tellili, B.; Souga, C.; Manai, K.

    2015-01-01

    Highlights: • Description of the theoretical method used by the ParShield program. • Description of the ParShield program. • Test and validation the ParShield program. - Abstract: This study aims to present a new computer program called ParShield which determines the neutron and gamma-ray shielding parameters. This program can calculate the total mass attenuation coefficients (μ t ), the effective atomic numbers (Z eff ) and the effective electron densities (N eff ) for gamma rays and it can also calculate the effective removal cross-sections (Σ R ) for fast neutrons for mixtures and compounds. The results obtained for the gamma rays by using ParShield were compared with the results calculated by the WinXcom program and the measured results. The obtained values of (Σ R ) were tested by comparing them with the measured results,the manually calculated results and with the results obtained by using MERCSFN program and an excellent agreement was found between them. The ParShield program can be used as a fast and effective tool to choose and compare the shielding materials, especially for the determination of (Z eff ) and (N eff ), there is no other programs in the literature which can calculate

  11. Calculation of the detection limits for radionuclides identified in gamma-ray spectra based on post-processing peak analysis results.

    Science.gov (United States)

    Korun, M; Vodenik, B; Zorko, B

    2018-03-01

    A new method for calculating the detection limits of gamma-ray spectrometry measurements is presented. The method is applicable for gamma-ray emitters, irrespective of the influences of the peaked background, the origin of the background and the overlap with other peaks. It offers the opportunity for multi-gamma-ray emitters to calculate the common detection limit, corresponding to more peaks. The detection limit is calculated by approximating the dependence of the uncertainty in the indication on its value with a second-order polynomial. In this approach the relation between the input quantities and the detection limit are described by an explicit expression and can be easy investigated. The detection limit is calculated from the data usually provided by the reports of peak-analyzing programs: the peak areas and their uncertainties. As a result, the need to use individual channel contents for calculating the detection limit is bypassed. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Site dose calculations for the INEEL/TMI-2 storage facility

    International Nuclear Information System (INIS)

    Jones, K.B.

    1997-01-01

    The U.S. Department of Energy (DOE) is licensing an independent spent-fuel storage installation (ISFSI) for the Three Mile Island unit 2 (TMI-2) core debris to be constructed at the Idaho Chemical Processing Plant (ICPP) site at the Idaho National Engineering and Environmental Laboratory (INEEL) using the NUHOMS spent-fuel storage system. This paper describes the site dose calculations, performed in support of the license application, that estimate exposures both on the site and for members of the public. These calculations are unusual for dry-storage facilities in that they must account for effluents from the system in addition to skyshine from the ISFSI. The purpose of the analysis was to demonstrate compliance with the 10 CFR 20 and 10 CFR 72.104 exposure limits

  13. Method to calculating an internal electromagnetic pulse generated in a system under gamma radiation effect; Metod rascheta vnutrennego ehlektromagnitnogo impul`sa, generiruemogo v sisteme pri vozdejstvii gamma-izlucheniya

    Energy Technology Data Exchange (ETDEWEB)

    Ogorodnikov, S N

    1994-12-31

    A method of calculating internal electromagnetic pulse, generated in the system under effect of gamma radiation is developed. Ratios for basic electron flux characteristics and components of electric and magnetic fields generated by gamma radiation, are indicated for a cylindrical cavity under gamma radiation effect on its surface. To illustrate this a case is considered when a single flux velocity component is present.

  14. Integrated technique for assessing environmental dose of radioactive waste storage installation

    Energy Technology Data Exchange (ETDEWEB)

    Bor-Jing Chang; Chien-Liang Shih; Ing-Jane Chen [Institute of Nuclear Energy Research, Lungtan, Taiwan (China); Ren-Dih Sheu; Shiang-Huei Jiang [National Tsing Hua University, Hsinchu, Taiwan (China); Shu-Jun Chang [Nuclear Science and Technology Association, Taiwan (China); Ruei-Ying Liao; Pei Yu; Chin-Yi Huang [Taiwan Power Company, Taipei, Taiwan (China)

    2000-05-01

    The ability to accurately predict exposure rates at large distances from a gamma radiation source is becoming increasingly important. This is because that the related regulation for the control of radiation levels in and around nuclear facilities becomes more stringent. Since the continuous increase of the radwaste storage capacity requirement on site, the requirement of a more realistic evaluation is very necessary. Those doses are usually at past time evaluated by QADCG/INER-2 code for direct dose and by SKYSHINE-III code for skyshine dose in which evaluation were over conservatively considered. This study is to update the evaluation code package accompanied with adequate methodology and to establish integrated analysis procedure. Thereafter, radiation doses can be accurately calculated in a reasonably conservative way. The purpose of the investigation is divided into three categories. First, SPECTRUM-506 is used instead of SPECTUM. Nuclide databases are enlarged from 100 up to 506. And the operation is ported to personal computer. Secondly, the QADCG/INER-3 code is developed to enhance the original QADCG/INER-2 code. The most important difference is the use of the geometric progression (GP) fitting function for the gamma-ray buildup factor. SKYSHINE-III code is replaced by McSKY and SKYDOSE codes. They are well benchmarked by using the Monte Carlo code MCNP and sensitive parameters are detailed investigated. Thirdly, the well developed analysis procedure is applicable for nuclear utility radwaste storage sites. Finally, the case studies were performed by using those packages to assess the radiological impact of utility radwaste storage site. The results are verified in detail by using Monte Carlo code MCNP and the results seems pretty consistent from both method. (author)

  15. Neutron and gamma ray calculation for Hiroshima-type atomic bomb

    Energy Technology Data Exchange (ETDEWEB)

    Hoshi, Masaharu; Endo, Satoru; Takada, Jun [Hiroshima Univ. (Japan). Research Inst. for Radiation Biology and Medicine; Iwatani, Kazuo; Oka, Takamitsu; Shizuma, Kiyoshi; Fujita, Shoichiro; Hasai, Hiromi

    1998-03-01

    We looked at the radiation dose of Hiroshima and Nagasaki atomic bomb again in 1986. We gave it the name of ``Dosimetry System 1986`` (DS86). We and other groups have measured the expose dose since 1986. Now, the difference between data of {sup 152}Eu and the calculation result on the basis of DS86 was found. To investigate the reason, we carried out the calculations of neutron transport and neutron absorption gamma ray for Hiroshima atomic bomb by MCNP3A and MCNP4A code. The problems caused by fast neutron {sup 32}P from sulfur in insulator of pole. To correct the difference, we investigated many models and found agreement of all data within 1 km. (S.Y.)

  16. Some Problems of Calculation and Design of High-Activity Gamma Units; Quelques aspects du calcul et de l'etablissement de projets d'installations puissantes emettrices de rayonnement gamma; Nekotorye voprosy rascheta i proektirovaniya moshchnykh gamma-ustanovok; Algunos aspectos del calculo y construccion de instalaciones de irradiacion gamma de elevada intensidad

    Energy Technology Data Exchange (ETDEWEB)

    Bibergal, A V; Leshchinsky, N I; Margulis, U Ya; Khrushev, V G [Academy of Sciences of the USSR, Moscow, Union of Soviet Socialist Republics (Russian Federation)

    1960-07-15

    The report describes the principal requirements for gamma units intended for various purposes. Several methods of calculating the dose fields for various forms of irradiators are given, as well as graphs, nomograms and formulae to estimate the amount of gamma-ray dose absorbed by the irradiated object from cobalt-60 and caesium-137. Some of the calculated data have been confirmed by experiment. The advantages of irradiators of various geometry employed in experimental and commercial units are discussed. The irradiation technique for various objects is analyzed and the optimum irradiation conditions (radiation utilization factor, dose field homogeneity, etc.) are discussed. Several rational shielding systems are suggested to simplify irradiation process and recharging, and to reduce the cost of design and operation. (author) [French] Resume Cette communication expose les conditions fondamentales auxquelles doivent satisfaire les installations emettrices de rayonnement gamma, destinees, a divers usages. Elle explique certaines methodes pour le calcul des champs recevant une dose determinee en fonction de la forme du dispositif d'irradiation; elle contient des courbes, des nomogrammes et des formules permettant de calculer la valeur de la dose de rayons gamma du cobalt-60 et du cesium-137 recue par l'echantillon soumis a l'irradiation. Certaines donnees etablies par le calcul ont ete confirmees experimentalement. On examine l'utilite d'employer des dispositifs d'irradiation de configurations variees dans des installations experimentales et des installations industrielles. Les auteurs examinent la technologie de l'irradiation de divers echantillons en vue de choisir les conditions optimales d'irradiation (facteur d'utilisation du rayonnement, uniformite du champ recevant une dose determinee, etc.). Quelques systemes rationnels de protection, simplifiant le processus d'irradiation, sont exposes ainsi que des methodes permettant un echange de charge dans les

  17. Calculation of the energy spectrum of atmospheric gamma-rays between 1 and 1000 MeV

    International Nuclear Information System (INIS)

    Martin, I.M.; Dutra, S.L.G.; Palmeira, R.A.R.

    The energy spectrum of atmospheric gamma-rays at 4 g/cm 2 has been calculated for cut-off rigidities of 4.5, 10 and 16 GV. The considered processes for the production of these gamma-rays were the π 0 decay plus the bremsstrahlung from primary, secondary like splash and re-entrant albedo electrons. The calculations indicated that the spectrum could be fitted to a power law in energy, with the exponential index varying from 1.1 in the energy range 1 - 10 MeV, to 1.4 in the energy range 10 - 200 MeV and 1.8 in the energy range 200 - 1000 MeV. These results are discussed [pt

  18. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  19. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  20. Study of {gamma}'s in Naiade; Etude des gamma de Naiade

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    Following a study of the gamma sources, the flux of gamma of different energies in the swimming pool is investigated. The biological dose can thus be obtained by calculation, and compared with the results given by photographic plates. The influence of photoneutrons is estimated by calculation, and research is being carried out on their influence on the thermal neutron flux curve on the axis of the uranium plate, with the plate emitting neutrons and with the plate protected by boral. (author) [French] Apres l'etude des sources de gamma, l'on etudie le flux de gamma de differentes energies dans la piscine. La dose biologique peut etre obtenue ainsi par le calcul et comparee avec les resultats donnes par les plaques photographiques. L'influence des photoneutrons est estimee par le calcul et l'on recherche leur influence sur la courbe de flux de neutrons thermiques sur l'axe de la plaque d'uranium, la plaque emettant des neutrons et la plaque protegee par du boral. (auteur)

  1. Laboratory calibrations of airborne gamma-ray spectrometers. Measurements and discussions of important parameters

    International Nuclear Information System (INIS)

    Korsbech, U.

    1994-02-01

    This report is the fourth of reports from The Department of Electrophysics covering measurement and interpretation of airborne gamma-spectrometry measurements. It describes different topics concerning the construction of a suitable calibration setup in the laboratory. The goal is to build a simple and cheap laboratory setup that can produce most of the gamma-ray data needed for an interpretation of spectra measured 50 to 120 m above ground level. A simple calibration setup has been build and tested. It may produce gamma-ray spectra similar to those measured in the air - from surface contamination with artificial nuclides and from 'bulk' natural radioactivity. It is possible to investigate the influence of the air above an aircraft carrying the detector (skyshine: scattering of gamma photons in the air above the detector). In order to reduce the influence of non-detected pile-up the count rates are kept low without reaching levels where the background spectra (to be subtracted) would cause unacceptable counting statistical fluctuations. Sources selected for the calibrations are heavy minerals sand (with thorium and uranium), potassium nitrate (with 40 K). These sources are 'bulk sources' of natural radioactivity. Cesium-137 has been selected as the basic artifical surface contamination nuclide. The report also discusses methods for comparing two spectra a priori assumed equal. Finally the properties of some materials that could be used as 'air-substitutes' in the calibration setup have been tested with respect to stability against moisture sorption. (au)

  2. Analytical calculations of the efficiency of gamma scintillators total efficiency for coaxial disk sources

    Energy Technology Data Exchange (ETDEWEB)

    Selim, Y S; Abbas, M I; Fawzy, M A [Physics Department, Faculty of Science, Alexandria University, Aleaxndria (Egypt)

    1997-12-31

    Total efficiency of clad right circular cylindrical Nal(TI) scintillation detector from a coaxial isotropic radiating circular disk source has been calculated by the of rigid mathematical expressions. Results were tabulated for various gamma energies. 2 figs., 5 tabs.

  3. Monte Carlo simulation of the Leksell Gamma Knife: I. Source modelling and calculations in homogeneous media

    Energy Technology Data Exchange (ETDEWEB)

    Moskvin, Vadim [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN (United States)]. E-mail: vmoskvin@iupui.edu; DesRosiers, Colleen; Papiez, Lech; Timmerman, Robert; Randall, Marcus; DesRosiers, Paul [Department of Radiation Oncology, Indiana University School of Medicine, Indianapolis, IN (United States)

    2002-06-21

    The Monte Carlo code PENELOPE has been used to simulate photon flux from the Leksell Gamma Knife, a precision method for treating intracranial lesions. Radiation from a single {sup 60}Co assembly traversing the collimator system was simulated, and phase space distributions at the output surface of the helmet for photons and electrons were calculated. The characteristics describing the emitted final beam were used to build a two-stage Monte Carlo simulation of irradiation of a target. A dose field inside a standard spherical polystyrene phantom, usually used for Gamma Knife dosimetry, has been computed and compared with experimental results, with calculations performed by other authors with the use of the EGS4 Monte Carlo code, and data provided by the treatment planning system Gamma Plan. Good agreement was found between these data and results of simulations in homogeneous media. Owing to this established accuracy, PENELOPE is suitable for simulating problems relevant to stereotactic radiosurgery. (author)

  4. Calculation of the gamma-dose rate from a continuously emitted plume

    International Nuclear Information System (INIS)

    Huebschmann, W.; Papadopoulos, D.

    1975-06-01

    A computer model is presented which calculates the long term gamma dose rate caused by the radioactive off-gas continuously emitted from a stack. The statistical distribution of the wind direction and velocity and of the stability categories is taken into account. The emitted activity, distributed in the atmosphere according to this statistics, is assumed to be concentrated at the mesh points of a three-dimensional grid. The grid spacing and the integration limits determine the accuracy as well as the computer time needed. When calculating the dose rate in a given wind direction, the contribution of the activity emitted into the neighbouring sectors is evaluated. This influence is demonstrated in the results, which are calculated with a error below 3% and compared to the dose rate distribution curves of the submersion model and the model developed by K.J. Vogt. (orig.) [de

  5. A semi-empirical approach to calculate gamma activities in environmental samples

    International Nuclear Information System (INIS)

    Palacios, D.; Barros, H.; Alfonso, J.; Perez, K.; Trujillo, M.; Losada, M.

    2006-01-01

    We propose a semi-empirical method to calculate radionuclide concentrations in environmental samples without the use of reference material and avoiding the typical complexity of Monte-Carlo codes. The calculation of total efficiencies was carried out from a relative efficiency curve (obtained from the gamma spectra data), and the geometric (simulated by Monte-Carlo), absorption, sample and intrinsic efficiencies at energies between 130 and 3000 keV. The absorption and sample efficiencies were determined from the mass absorption coefficients, obtained by the web program XCOM. Deviations between computed results and measured efficiencies for the RGTh-1 reference material are mostly within 10%. Radionuclide activities in marine sediment samples calculated by the proposed method and by the experimental relative method were in satisfactory agreement. The developed method can be used for routine environmental monitoring when efficiency uncertainties of 10% can be sufficient.(Author)

  6. Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4

    International Nuclear Information System (INIS)

    Baur, A.; Dupont, C.; Totth, B.

    1978-01-01

    The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks

  7. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  8. Radiative decay of the eta-, eta'-mesons in the nonlocal quark model. [eta(eta'). --> gamma gamma. ; eta. -->. pi. /sup +/. pi. /sup -/. gamma. ; eta. -->. pi. /sup 0/2. gamma. ; eta'. -->. rho/sup 0/. gamma. ; eta'. -->. omega gamma. ;. pi. /sup 0/. -->. gamma. e/sup +/e/sup -/; eta(eta'). -->. gamma mu. /sup +/. mu. /sup -/

    Energy Technology Data Exchange (ETDEWEB)

    Efimov, G V; Ivanov, M A; Nogovitsyn, E A [Joint Inst. for Nuclear Research, Dubna (USSR)

    1981-07-01

    P..--> gamma gamma.. (P=..pi../sup 0/, eta, eta'), eta..--> pi../sup +/..pi../sup -/..gamma.., eta..--> pi../sup 0/..gamma gamma.., eta/sup 1/..-->..V..gamma.. (V=rho/sup 0/, ..omega..), p..--> gamma..l/sup +/l/sup -/ (p=..pi../sup 0/, eta, eta') radiation decays are studied for testing the applicability of the non-local quark model for description of the experimental data. The Feynman diagrams of these decays are presented, values of the widths of the Veta..--> gamma gamma.., eta..--> pi../sup +/..pi../sup -/..gamma.., eta..--> pi../sup 0/..gamma gamma.., eta'..--> gamma gamma.., eta'..-->..rho/sup 0/..gamma.., eta'..--> omega gamma.. decays are calculated and given in the form of a table. Calculations are carried out for two values of the eta eta'-crossing angle: THETA=-11 deg and -18 deg. Values of invariant amplitudes of these decays are determined for ..pi../sup 0/..--> gamma..e/sup +/e/sup -/, eta..--> gamma mu../sup +/..mu../sup -/, eta'..--> gamma mu../sup +/..mu../sup -/ decays at THETA=-11 deg and -18 deg. The best agreement with the experimental data is noted to take place at THETA=-11 deg, the determined width of the eta..--> pi../sup 0/..gamma gamma.. decays is underestimated as compared with the experimental one.

  9. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 2. Attenuation of gamma rays. An example of shield's thickness calculation

    International Nuclear Information System (INIS)

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)

  10. An analytical model to calculate absorbed fractions for internal dosimetry with alpha, beta and gamma emitters

    Directory of Open Access Journals (Sweden)

    Ernesto Amato

    2014-03-01

    Full Text Available We developed a general model for the calculation of absorbed fractions in ellipsoidal volumes of soft tissue uniformly filled with alpha, beta and gamma emitting radionuclides. The approach exploited Monte Carlo simulations with the Geant4 code to determine absorbed fractions in ellipsoids characterized by a wide range of dimensions and ellipticities, for monoenergetic emissions of each radiation type. The so-obtained absorbed fractions were put in an analytical relationship with the 'generalized radius', calculated as 3V/S, where V is the ellipsoid volume and S its surface. Radiation-specific parametric functions were obtained in order to calculate the absorbed fraction of a given radiation in a generic ellipsoidal volume. The dose from a generic radionuclide can be calculated through a process of summation and integration over the whole radionuclide emission spectrum, profitably implemented in an electronic spreadsheet. We compared the results of our analytical calculation approach with those obtained from the OLINDA/EXM computer software, finding a good agreement in a wide range of sphere radii, for the high-energy pure beta emitter 90Y, the commonly employed beta-gamma emitter 131I, and the pure alpha emitter 213Po. The generality of our approach makes it useful an easy to implement in clinical dosimetry calculations as well as in radiation safety estimations when doses from internal radionuclide uptake are to be taken into account.

  11. Measurement and calculation of secondary gamma rays resulting from exposure of Fe, Pb, and H/sub 2/O to the ARERR-1 spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Makarious, A.S.; Ford, W.E. III; Turnbull, K.R.

    1977-08-01

    Integral experiments were performed to measure the angular distribution of secondary gamma rays produced when various thicknesses of Fe, Pb, and H/sub 2/O samples were exposed to bare and to B/sub 4/C-filtered neutron beams from the Research Reactor of Egypt. For selected experiments, multigroup coupled neutron-gamma cross sections and a discrete ordinates transport theory code (DOT4PI-M) were used to calculate the secondary gamma rays and the transport of primary gamma rays. Integral comparisons between the calculated and measured spectra were favorable. Graphical comparisons of the measured flux for various angles of incidence of the neutron beams on the samples, for various angles of exit on the transmitted side of the samples, and for various sample thicknesses are shown. The comparisons show that the angular distribution of secondary gamma rays for the three materials changes slightly with a change in the angle of beam incident on the sample, but increasing the angle between the normal to the sample and the detector by 60/sup 0/ decreases the measured secondary gamma-ray flux up to a factor of two. An investigation was made to determine the consequences of using single scatter Compton theory versus using discrete ordinates transport calculations to estimate the primary gamma-ray contribution to the measured photon spectra.

  12. A fence line noble gas monitoring system for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Grasty, R.L.; Hovgaard, J.; LaMarre, J.R

    2001-07-01

    A noble gas monitoring system has been installed at Ontario Power Generations' Pickering Nuclear Generating Station (PNGS) near Toronto, Canada. This monitoring system allows a direct measure of air kerma from external radiation instead of calculating this based on plant emission data and meteorological models. This has resulted in a reduction in the reported effective dose from external radiation by a factor of at least ten. The system consists of nine self-contained units, each with a 7.6 cm x 7.6 cm (3 inch x 3 inch) NaI(Tl) detector that is calibrated for air kerma. The 512-channel gamma ray spectral information is downloaded daily from each unit to a central computer where the data are stored and processed. A spectral stripping procedure is used to remove natural background variations from the spectral windows used to monitor xenon-133 ({sup 133}Xe), xenon-135 ({sup 135}Xe), argon-41 ({sup 41}Ar), and skyshine radiation from the use of radiography sources. Typical monthly minimum detection limits in air kerma are 0.3 nGy for {sup 133}Xe, 0.7 nGy for {sup 135}Xe, 3 nGy for {sup 41}Ar and 2 nGy for skyshine radiation. Based on 9 months of continuous operation, the annualised air kerma due to {sup 133}Xe, {sup 135}Xe and {sup 41}Ar and skyshine radiation were 7 nGy, 8 nGy, 26 nGy and 107 nGy respectively. (author)

  13. Calculating concentration of inhaled radiolabeled particles from external gamma counting: External counting efficiency and attenuation coefficient of thorax

    International Nuclear Information System (INIS)

    Langenback, E.G.; Foster, W.M.; Bergofsky, E.H.

    1989-01-01

    We determined the overall external counting efficiency of radiolabeled particles deposited in the sheep lung. This efficiency permits the noninvasive calculation of the number of particles and microcuries from gamma-scintillation lung images of the live sheep. Additionally, we have calculated the attenuation of gamma radiation (120 keV) by the posterior chest wall and the gamma-scintillation camera collection efficiency of radiation emitted from the lung. Four methods were employed in our experiments: (1) by light microscopic counting of discrete carbonized polystyrene particles with a count median diameter (CMD) of 2.85 microns and tagged with cobalt-57, we delineated a linear relationship between the number of particles and the emitted counts per minute (cpm) detected by well scintillation counting; (2) from this conversion relationship we determined the number of particles inhaled and deposited in the lungs by scintillation counting fragments of dissected lung at autopsy; (3) we defined a linear association between the number of particles or microcuries contained in the lung and the emitted radiation as cpm detected by a gamma scintillation camera in the live sheep prior to autopsy; and (4) we compared the emitted radiation from the lungs of the live sheep to that of whole excised lungs in order to calculate the attenuation coefficient (ac) of the chest wall. The mean external counting efficiency was 4.00 X 10(4) particles/cpm (5.1 X 10(-3) microCi/cpm), the camera collection efficiency was 1 cpm/10(4) disintegrations per minute (dpm), and the ac had a mean of 0.178/cm. The external counting efficiency remained relatively constant over a range of particles and microcuries, permitting a more general use of this ratio to estimate number of particles or microcuries depositing after inhalation in a large mammalian lung if a similarly collimated gamma camera system is used

  14. Neutron and gamma ray streaming calculations for the ETF neutral beam injectors

    International Nuclear Information System (INIS)

    Lillie, R.A.; Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1981-02-01

    Two-dimensional radiation transport methods have been used to estimate the effects of neutron and gamma ray streaming on the performance of the Engineering Test Facility (ETF) neutral beam injectors. The calculations take into account the spatial, angular, and spectral distributions of the radiation entering the injector duct. The instantaneous nuclear heating rate averaged over the length of the cryopumping panel in the injector is 7.5 x 10 -3 MW/m 3 which implies a total heat load of 2.2 x 10 -4 MW. The instantaneous dose rate to the ion gun insulators was estimated to be 3200 rad/s. The radial dependence of the instantaneous dose equivalent rate in the neutral beam injector duct shield was also calculated

  15. Atmospheric gamma-ray observation with the BETS detectorfor calibrating atmospheric neutrino flux calculations

    CERN Document Server

    Kasahara, K.; Torii, S.; Tamura, T.; Tateyama, N.; Yoshida, K.; Yamagami, T.; Saito, Y.; Nishimura, J.; Murakami, H.; Kobayashi, T.; Komori, Y.; Honda, M.; Ohuchi, T.; Midorikawa, S.; Yuda, T.

    2002-01-01

    We observed atmospheric gamma-rays around 10 GeV at balloon altitudes (15~25 km) and at a mountain (2770 m a.s.l). The observed results were compared with Monte Carlo calculations to find that an interaction model (Lund Fritiof1.6) used in an old neutrino flux calculation was not good enough for describing the observed values. In stead, we found that two other nuclear interaction models, Lund Fritiof7.02 and dpmjet3.03, gave much better agreement with the observations. Our data will serve for examining nuclear interaction models and for deriving a reliable absolute atmospheric neutrino flux in the GeV region.

  16. Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

    Science.gov (United States)

    Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

    2018-05-01

    Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

  17. FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

    International Nuclear Information System (INIS)

    Stamatelatos, M.G.; England, T.R.

    1977-05-01

    FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

  18. Approximate techniques for calculating gamma ray dose rates in nuclear power plants

    International Nuclear Information System (INIS)

    Lahti, G.P.

    1986-01-01

    Although today's computers have made three-dimensional discrete ordinates transport codes a virtual reality, there is still a need for approximate techniques for estimating radiation environments. This paper discusses techniques for calculating gamma ray dose rates in nuclear power plants where Compton scattering is the dominant attenuation mechanism. The buildup factor method is reviewed; its use and misuse are discussed. Several useful rules-of-thumb are developed. The paper emphasizes the need for understanding the fundamental physics and draws heavily on the old, classic references

  19. Dosimetric calculations by Monte Carlo for treatments of radiosurgery with the Leksell Gamma Knife, homogeneous and non homogeneous cases

    International Nuclear Information System (INIS)

    Rojas C, E.L.; Lallena R, A.M.

    2004-01-01

    In this work dose profiles are calculated that are obtained modeling treatments of radiosurgery with the Leksell Gamma Knife. This was made with the simulation code Monte Carlo Penelope for an homogeneous mannequin and one not homogeneous. Its were carried out calculations with the irradiation focus coinciding with the center of the mannequin as in near areas to the bone interface. Each one of the calculations one carries out for the 4 skull treatment that it includes the Gamma Knife and using a model simplified of their 201 sources of 60 Co. It was found that the dose profiles differ of the order of 2% when the isocenter coincides with the center of the mannequin and they ascend to near 5% when the isocenter moves toward the skull. (Author)

  20. Applying ISO 11929:2010 Standard to detection limit calculation in least-squares based multi-nuclide gamma-ray spectrum evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kanisch, G., E-mail: guenter.kanisch@hanse.net

    2017-05-21

    The concepts of ISO 11929 (2010) are applied to evaluation of radionuclide activities from more complex multi-nuclide gamma-ray spectra. From net peak areas estimated by peak fitting, activities and their standard uncertainties are calculated by weighted linear least-squares method with an additional step, where uncertainties of the design matrix elements are taken into account. A numerical treatment of the standard's uncertainty function, based on ISO 11929 Annex C.5, leads to a procedure for deriving decision threshold and detection limit values. The methods shown allow resolving interferences between radionuclide activities also in case of calculating detection limits where they can improve the latter by including more than one gamma line per radionuclide. The co'mmon single nuclide weighted mean is extended to an interference-corrected (generalized) weighted mean, which, combined with the least-squares method, allows faster detection limit calculations. In addition, a new grouped uncertainty budget was inferred, which for each radionuclide gives uncertainty budgets from seven main variables, such as net count rates, peak efficiencies, gamma emission intensities and others; grouping refers to summation over lists of peaks per radionuclide.

  1. Effects of bone- and air-tissue inhomogeneities on the dose distributions of the Leksell Gamma Knife (registered) calculated with PENELOPE

    International Nuclear Information System (INIS)

    Al-Dweri, Feras M O; Rojas, E Leticia; Lallena, Antonio M

    2005-01-01

    Monte Carlo simulation with PENELOPE (version 2003) is applied to calculate Leksell Gamma Knife (registered) dose distributions for heterogeneous phantoms. The usual spherical water phantom is modified with a spherical bone shell simulating the skull and an air-filled cube simulating the frontal or maxillary sinuses. Different simulations of the 201 source configuration of the Gamma Knife have been carried out with a simplified model of the geometry of the source channel of the Gamma Knife recently tested for both single source and multisource configurations. The dose distributions determined for heterogeneous phantoms including the bone- and/or air-tissue interfaces show non-negligible differences with respect to those calculated for a homogeneous one, mainly when the Gamma Knife isocentre approaches the separation surfaces. Our findings confirm an important underdosage (∼10%) nearby the air-tissue interface, in accordance with previous results obtained with the PENELOPE code with a procedure different from ours. On the other hand, the presence of the spherical shell simulating the skull produces a few per cent underdosage at the isocentre wherever it is situated

  2. Assessment of neutron skyshine near unmodified Accumulator Debuncher storage rings under Mu2e operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cossairt, J.Donald; /Fermilab

    2010-12-01

    Preliminary plans for providing the proton beam needed by the proposed Mu2e experiment at Fermilab will require the transport of 8 GeV protons to the Accumulator/Debuncher where they be processed into an intensity and time structure useful for the experiment. The intensities involved are far greater that those encountered with antiprotons of the same kinetic energy in the same beam enclosures under Tevatron Collider operational conditions, the operating parameters for which the physical facilities of the Antiproton Source were designed. This note explores some important ramifications of the proposed operation for radiation safety and demonstrates the need for extensive modifications of significant portions of the shielding of the Accumulator Debuncher storage rings; notably that underneath the AP Service Buildings AP10, AP30, and AP50. While existing shielding is adequate for the current operating mode of the Accumulator/Debuncher as part of the Antiproton Source used in the Tevatron Collider program, without significant modifications of the shielding configuration in the Accumulator/Debuncher region and/or beam loss control systems far more effective than seen in most applications at Fermilab, the proposed operational mode for Mu2e is not viable for the following reasons: 1. Due to skyshine alone, under normal operational conditions large areas of the Fermilab site would be exposed to unacceptable levels of radiation where most of the Laboratory workforce and some members of the general public who regularly visit Fermilab would receive measurable doses annually, contrary to workforce, public, and DOE expectations concerning the As Low as Reasonably Achievable (ALARA) principle. 2. Under normal operational conditions, a sizeable region of the Fermilab site would also require fencing due to skyshine. The size of the areas involved would likely invite public inquiry about the significant and visible enlargement of Fermilab's posted radiological areas. 3. There

  3. Calculation of calibration factors and layout criteria for gamma scanning of waste drums from nuclear plants

    International Nuclear Information System (INIS)

    Inder Schmitten, W.; Sohnius, B.; Wehner, E.

    1990-01-01

    This paper present a procedure to calculate calibration factors for converting the measured gamma rate of waste drums into activity content and a layout and free release measurement criterion for waste drums. A computer program is developed that simulates drum scanning technique, which calculates calibration factors and eliminates laborious experimental measurements. The calculated calibration factors exhibit good agreement with experimentally determined values. By checking the calculated calibration factors for trial equipment layouts (including the waste drum and the scanning facility) using the layout and free release measurement criterion, a layout can be achieved that clearly determines whether there can be free release of a waste drum

  4. Frontal midline theta rhythm and gamma power changes during focused attention on mental calculation: an MEG beamformer analysis

    Directory of Open Access Journals (Sweden)

    Ryouhei eIshii

    2014-06-01

    Full Text Available Frontal midline theta rhythm (Fmθ appears widely distributed over medial prefrontal areas in EEG recordings, indicating focused attention. Although mental calculation is often used as an attention-demanding task, little has been reported on calculation-related activation in Fmθ experiments. In this study we used spatially filtered MEG and permutation analysis to precisely localize cortical generators of the magnetic counterpart of Fmθ, as well as other sources of oscillatory activity associated with mental calculation processing (i.e., arithmetic subtraction. Our results confirmed and extended earlier EEG/MEG studies indicating that Fmθ during mental calculation is generated in the dorsal anterior cingulate and adjacent medial prefrontal cortex. Mental subtraction was also associated with gamma event-related synchronization, as an index of activation, in right parietal regions subserving basic numerical processing and number-based spatial attention. Gamma event-related desynchronization appeared in the right lateral prefrontal cortex, likely representing a mechanism to interrupt neural activity that can interfere with the ongoing cognitive task.

  5. Bhabha Atomic Research Centre, Bombay

    International Nuclear Information System (INIS)

    Swarup, J.; Ganguly, A.K.

    1977-01-01

    The paper reports the preliminary results obtained on the sky-shine spectra from a 650 Ci 60 Co source located at the center of a gamma irradiation field of radius 90 m fenced by a stone wall of thickness approximately 75 cm and height 3.66 m. The source is in the form of a small pellet. The height of the source when raised for irradiation is 1.2 m above ground level and it is shielded on top by a lead cylinder of 10 cm diameter and 25 cm length. Thus, only the scattered radiation can reach the ground level beyond the fencing wall. There is a field of 100 mR/hr on the inner side and 2 mR/hr on the outer side of the wall with the source raised. Experiments are carried out for the measurement of sky-shine with a well-shielded NaI detector assembly coupled to a 400-channel analyzer. The detector is placed 55 cm above ground looking vertically up through a lead collimator of diameter 12 mm (or 20 mm) at distances from 150 to 325 m away from the source. Energy calibrations of the spectra have been carried out before and after each experiment using standard sources of gamma-energy ranging from 60 to 662 keV. It is found that the spectrum extends up to 400 keV with a pronounced peak at 72 keV for all the distances. There is no evidence of the presence of primary gamma-photons in the spectra. Total counts under the sky-shine are observed to follow an exponential decline with distance, with a slope of -0.50 +- 0.02 for both the collimators used. The ratio of peak counts (72 keV) to total sky-shine is 0.24 +- 0.02 for both the collimators. Also, the nature and intensity of the spectra remain unchanged when the lead shield around the detector is provided with an internal lining of 2.5 cm thick aluminium

  6. Calculation of neutron and gamma-ray energy spectra in liquid air and liquid nitrogen due to 14-MeV neutron and californium-252 sources

    International Nuclear Information System (INIS)

    Straker, E.A.; Gritzner, M.L.; Harris, L. Jr.

    1978-01-01

    Calculations of neutron and gamma-ray fluences from 14-MeV neutron and 252 Cf sources in liquid air and liquid nitrogen have been performed. These calculations were made specifically for comparison with experimental data measured at Stohl, Federal Republic of Germany. The discrete-ordinates method was utilized with neutron and gamma-ray cross sections from ENDF/B-IV. One-dimensional calculational models were developed for the sources and tank. Limited comparisons are made with experimental data

  7. The effect of gamma-ray transport on afterheat calculations for accident analysis

    International Nuclear Information System (INIS)

    Reyes, S.; Latkowski, J.F.; Sanz, J.

    2000-01-01

    Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed

  8. Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tarchalski, M.; Pytel, K.; Wroblewska, M.; Marcinkowska, Z.; Boettcher, A.; Prokopowicz, R. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Sireta, P.; Gonnier, C.; Bignan, G. [CEA, DEN, Reactor Studies Department, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Fourmentel, D.; Barbot, L.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C.; Brun, J. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Jagielski, J. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Institute of Electronic Materials Technolgy, Wolczynska 133, 01-919 Warszawa (Poland); Luks, A. [Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland)

    2015-07-01

    Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from

  9. Dose calculation method with 60-cobalt gamma rays in total body irradiation

    International Nuclear Information System (INIS)

    Scaff, Luiz Alberto Malaguti

    2001-01-01

    Physical factors associated to total body irradiation using 60 Co gamma rays beams, were studied in order to develop a calculation method of the dose distribution that could be reproduced in any radiotherapy center with good precision. The method is based on considering total body irradiation as a large and irregular field with heterogeneities. To calculate doses, or doses rates, of each area of interest (head, thorax, thigh, etc.), scattered radiation is determined. It was observed that if dismagnified fields were considered to calculate the scattered radiation, the resulting values could be applied on a projection to the real size to obtain the values for dose rate calculations. In a parallel work it was determined the variation of the dose rate in the air, for the distance of treatment, and for points out of the central axis. This confirm that the use of the inverse square law is not valid. An attenuation curve for a broad beam was also determined in order to allow the use of absorbers. In this work all the adapted formulas for dose rate calculations in several areas of the body are described, as well time/dose templates sheets for total body irradiation. The in vivo dosimetry, proved that either experimental or calculated dose rate values (achieved by the proposed method), did not have significant discrepancies. (author)

  10. A comparison of semiconductor gamma spectrometric analysis using the peak net area calculations and the whole spectrum processing

    International Nuclear Information System (INIS)

    Krnac, S.; Koskelo, M.; Venkatamaran, R.

    1998-01-01

    This study was conducted to compare the results of gamma spectrometric analysis using the Scaling Confirmatory Factor Analysis (SCFA) method to that of Genie2K, which uses a more traditional method. Gamma ray spectra had had been acquired for several gamma standard sources, all of which except Co-57 and Eu-152 being single gamma ray emitting nuclides. These standard sources spanned the energy range from 60 keV (Am-241) to 1116 keV (Zn-65). The standard sources were counted at 3 different geometries at 3 different geometries, with source-detector distances of 0, 5, and 15 cm. Using single gamma ray spectra collected at a given counting geometry, and the certificate file, an efficiency calibration was created for that geometry. Three different test spectra, one for each counting geometry, had been created by combining several of the standard source spectra. The efficiency calibrations created for the 3 geometries were loaded into the respective spectrum files. Each test spectrum was analyzed using the standard Genie2K engines; Peak locate, Peak search, Interactive peak fit, Background subs-traction, Efficiency correction, and Nuclide Identification with interference analysis. The results of the various calculation steps were reported. In all 3 test cases, the SCFA method identified all the nuclides correctly. The K-40 activities calculated by the SCFA method were reasonably close to that from Genie2K analysis. In general, the quantitative results of the SCFA method were impressive in all 3 cases. On a positive note, the SCFA method did identify low yield gamma lines in Eu-152, which were not identified by the Genie2K analysis. This substantiates claim that the SCFA is more sensitive than the traditional method of spectrum analysis. (authors)

  11. A verification calculation of drum and pulley overhead travelling crane on gamma irradiator ISG-500

    International Nuclear Information System (INIS)

    Syamsurrijal Ramdja; Ari Satmoko; Sutomo Budihardjo

    2010-01-01

    It has been verified the calculation of drum and pulleys on cranes as facility the gamma irradiator ISG-500. Drum is a device for rolling steel ropes while the pulley is a circular pieces called disks, and both of which are made from metal or non-metal to transmit motion and force. It has been verified the calculation of forces on the drum, drum diameter and length, and pressuring force occurred on the drums. Likewise to the pulley, the pulley diameter calculations verification, size of disc and shaft power pulleys. From the verification results, it will be obtained whether the data drums and pulley device are safe or not safe to be used. (author)

  12. Calculation of “LS-curves” for coincidence summing corrections in gamma ray spectrometry

    Science.gov (United States)

    Vidmar, Tim; Korun, Matjaž

    2006-01-01

    When coincidence summing correction factors for extended samples are calculated in gamma-ray spectrometry from full-energy-peak and total efficiencies, their variation over the sample volume needs to be considered. In other words, the correction factors cannot be computed as if the sample were a point source. A method developed by Blaauw and Gelsema takes the variation of the efficiencies over the sample volume into account. It introduces the so-called LS-curve in the calibration procedure and only requires the preparation of a single standard for each sample geometry. We propose to replace the standard preparation by calculation and we show that the LS-curves resulting from our method yield coincidence summing correction factors that are consistent with the LS values obtained from experimental data.

  13. Activation of the JET vacuum vessel: a comparison of calculated with measured gamma-radiation fluxes and dose rates

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Sadler, G.; Avery, A.; Verschuur, K.A.

    1988-01-01

    The gamma-radiation dose-rates inside the JET vacuum vessel due to induced radioactivity were measured at intervals throughout the 1986 period of operation, and the decay gamma energy spectrum was measured during the subsequent lengthy shutdown. The dose-rates were found to be in good agreement with values calculated using the neutron yield records compiled from the time-resolved neutron yield monitor responses for individual discharges. This result provides strong support for the reliability of the neutron yield monitor calibration. (author)

  14. Calculation of reasonable exemption levels for surface contamination by measuring overall gamma ray

    International Nuclear Information System (INIS)

    Ogino, Haruyuki; Hattori, Takatoshi

    2008-01-01

    The present regulation on surface contamination [Bq/cm 2 ] is determined from a simple radiological model for the most hazardous radionuclides (Pu-239 for alpha emitters and Sr-90 for beta emitters) and its extremely conservative model is applied for all other alpha and beta emitters. In this study, reasonable exemption levels for surface contamination are calculated for each radionuclide by adopting an original radiological dose evaluation method for surface contamination that can be applied in radiation safety, transport safety and waste safety. Furthermore, a new concept of judging the exemption by estimating the overall contamination [Bq] on the objects from the measurement of gamma ray has been designed and a reasonable value was derived. We conclude that the overall exemption levels obtained by gamma ray measurement can be one order smaller than those obtained by the conventional method for some radionuclides, such as Mn-54, Co-60, Nb-94, Cs-134, Cs-137, Eu-152 and Eu-154. (author)

  15. Gamma-Ray Emission Spectra as a Constraint on Calculations of 234,236,238U Neutron-Capture Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kawano, Toshihiko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Baramsai, Bayarbadrakh [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jandel, Marian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vieira, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilhelmy, Jerry B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Becker, John A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Krticka, Milan [Charles Univ., Prague (Czech Republic)

    2015-05-28

    Neutron capture cross sections in the “continuum” region (>≈1 keV) and gamma-emission spectra are of importance to basic science and many applied fields. Careful measurements have been made on most common stable nuclides, but physicists must rely on calculations (or “surrogate” reactions) for rare or unstable nuclides. Calculations must be benchmarked against measurements (cross sections, gamma-ray spectra, and <Γγ>). Gamma-ray spectrum measurements from resolved resonances were made with 1 - 2 mg/cm2 thick targets; cross sections at >1 keV were measured using thicker targets. The results show that the shape of capture cross section vs neutron energy is not sensitive to the form of the strength function (although the magnitude is); the generalized Lorentzian E1 strength function is not sufficient to describe the shape of observed gamma-ray spectra; MGLO + “Oslo M1” parameters produces quantitative agreement with the measured 238U(n,γ) cross section; additional strength at low energies (~ 3 MeV) -- likely M1-- is required; and careful study of complementary results on low-lying giant resonance strength is needed to consistently describe observations.

  16. GammaWorkshops Proceedings

    International Nuclear Information System (INIS)

    Ramebaeck, H.; Straelberg, E.; Klemola, S.; Nielsen, Sven P.; Palsson, S.E.

    2012-01-01

    Due to a sparse interaction during the last years between practioners in gamma ray spectrometry in the Nordic countries, a NKS activity was started in 2009. This GammaSem was focused on seminars relevant to gamma spectrometry. A follow up seminar was held in 2010. As an outcome of these activities it was suggested that the 2011 meeting should be focused on practical issues, e.g. different corrections needed in gamma spectrometric measurements. This three day's meeting, GammaWorkshops, was held in September at Risoe-DTU. Experts on different topics relevant for gamma spectrometric measurements were invited to the GammaWorkshops. The topics included efficiency transfer, true coincidence summing corrections, self-attenuation corrections, measurement of natural radionuclides (natural decay series), combined measurement uncertainty calculations, and detection limits. These topics covered both lectures and practical sessions. The practical sessions included demonstrations of tools for e.g. corrections and calculations of the above meantioned topics. (Author)

  17. GammaWorkshops Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Ramebaeck, H. (ed.) (Swedish Defence Research Agency (Sweden)); Straalberg, E. (Institute for Energy Technology, Kjeller (Norway)); Klemola, S. (Radiation and Nuclear Safety Authority, STUK (Finland)); Nielsen, Sven P. (Technical Univ. of Denmark. Risoe National Lab. for Sustainable Energy, Roskilde (Denmark)); Palsson, S.E. (Icelandic Radiation Safety Authority (Iceland))

    2012-01-15

    Due to a sparse interaction during the last years between practioners in gamma ray spectrometry in the Nordic countries, a NKS activity was started in 2009. This GammaSem was focused on seminars relevant to gamma spectrometry. A follow up seminar was held in 2010. As an outcome of these activities it was suggested that the 2011 meeting should be focused on practical issues, e.g. different corrections needed in gamma spectrometric measurements. This three day's meeting, GammaWorkshops, was held in September at Risoe-DTU. Experts on different topics relevant for gamma spectrometric measurements were invited to the GammaWorkshops. The topics included efficiency transfer, true coincidence summing corrections, self-attenuation corrections, measurement of natural radionuclides (natural decay series), combined measurement uncertainty calculations, and detection limits. These topics covered both lectures and practical sessions. The practical sessions included demonstrations of tools for e.g. corrections and calculations of the above meantioned topics. (Author)

  18. Radiation transport simulation in gamma irradiator systems using E G S 4 Monte Carlo code and dose mapping calculations based on point kernel technique

    International Nuclear Information System (INIS)

    Raisali, G.R.

    1992-01-01

    A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained

  19. A COMPARISON OF MEASURED AND CALCULATED GAMMA RAY ATTENUATION FOR A COMMON COUNTING GEOMETRY

    International Nuclear Information System (INIS)

    Gaylord, R F

    2004-01-01

    In order to perform quantitative gamma spectroscopy, it is necessary to know the sample-specific detection efficiency for photons as a function of energy. The detection efficiency, along with the branching ratio for the isotope and gamma ray of interest, is used to convert observed counts/second to actual disintegrations/second, and, hence, has a large effect on the accuracy of the measurement. In cases where the geometry of the source is simple and reproducible, such as a point source, small vial of solid, or jar of liquid, geometry-specific standards may be counted to determine the detection efficiency. In cases where the samples are large, irregular, or unique, this method generally cannot be used. For example, it is impossible to obtain a NIST-traceable standard glovebox or 55-gallon drum. In these cases, a combination of measured absolute detector efficiency and calculated sample-specific correction factors is commonly used. The correction factors may be calculated via Monte Carlo simulation of the item (the method used by Canberra's ISOCS system), or via semi-empirical calculation of matrix and container attenuations based on the thickness and composition of the container and radioactive matrix (ISOTOPIC by EG and G Ortec uses this method). The accuracy of these correction factors for specific geometries is often of vital interest when assessing the quality of gamma spectroscopy data. During the Building 251 Risk-Reduction Project, over 100 samples of high activity actinides will be characterized via gamma spectroscopy, typically without removing the material from the current storage containers. Most of the radioactive materials in B-251 are stored in cylindrical stainless steel canisters (called USV containers, after the Underground Storage Vaults they are commonly stored in), 13 cm in diameter, by 28 cm high, with walls that are 1.8 mm thick. While the actual samples have a variety of configurations inside the USV container, a very common configuration is

  20. Transport calculations of. gamma. -ray flux density and dose rate about implantable californium-252 sources

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, A; Lin, B I [Cincinnati Univ., Ohio (USA). Dept. of Chemical and Nuclear Engineering; Windham, J P; Kereiakes, J G

    1976-07-01

    ..gamma.. flux density and dose rate distributions have been calculated about implantable californium-252 sources for an infinite tissue medium. Point source flux densities as a function of energy and position were obtained from a discrete-ordinates calculation, and the flux densities were multiplied by their corresponding kerma factors and added to obtain point source dose rates. The point dose rates were integrated over the line source to obtain line dose rates. Container attenuation was accounted for by evaluating the point dose rate as a function of platinum thickness. Both primary and secondary flux densities and dose rates are presented. The agreement with an independent Monte Carlo calculation was excellent. The data presented should be useful for the design of new source configurations.

  1. Improvement of gamma-ray Sn transport calculations including coherent and incoherent scatterings and secondary sources of bremsstrahlung and fluorescence: Determination of gamma-ray buildup factors

    International Nuclear Information System (INIS)

    Kitsos, S.; Diop, C.M.; Assad, A.; Nimal, J.C.; Ridoux, P.

    1996-01-01

    Improvements of gamma-ray transport calculations in S n codes aim at taking into account the bound-electron effect of Compton scattering (incoherent), coherent scattering (Rayleigh), and secondary sources of bremsstrahlung and fluorescence. A computation scheme was developed to take into account these phenomena by modifying the angular and energy transfer matrices, and no modification in the transport code has been made. The incoherent and coherent scatterings as well as the fluorescence sources can be strictly treated by the transfer matrix change. For bremsstrahlung sources, this is possible if one can neglect the charged particles path as they pass through the matter (electrons and positrons) and is applicable for the energy range of interest for us (below 10 MeV). These improvements have been reported on the kernel attenuation codes by the calculation of new buildup factors. The gamma-ray buildup factors have been carried out for 25 natural elements up to 30 mean free paths in the energy range between 15 keV and 10 MeV

  2. The calculation of external gamma-ray doses from airborne and deposited radionuclides in the environmental code NECTAR

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1982-02-01

    A computer program has been developed for the rapid evaluation of external gamma-ray doses from airborne and deposited radionuclide mixtures. Based on a gaussian dispersion model, the program calculates the dose at any position, including points high above ground level or upwind of the source. Meteorological frequency data for wind speed, direction, atmospheric stability and rainfall are fully taken into account. The calculational model assumes that the ground surface is perfectly flat and that gamma-ray paths are entirely in air; the possible errors caused by these and other assumptions are discussed, with suggested correction factors. The program applies various criteria to determine the best approximation or numerical integration method for each target point; execution times (on an IBM 370 machine) thus vary from less than 0.01s to about 0.3s per target point for a single weather category. The program has been incorporated in the environmental release program NECTAR. (author)

  3. FTR europia gamma heating

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1975-01-01

    Calculated and experimental gamma heating rates of europia in the Engineering Mockup Critical Assembly (EMC) were correlated. A calculated to experimental (C/E) ratio of 1.086 was established in validating the theoretical approach and computational technique applied in the calculations. Gamma heat deposition rates in the FTR with Eu 2 O 3 control absorbers were determined from three-dimensional calculations. Maximum gamma heating was found to occur near the tip of a half-inserted row 5 control rod assembly--12.8 watts/gm of europia. Gamma heating profiles were established for a single half-inserted europia absorber assembly. Local heat peaking was found not to alter significantly heating rates computed in the FTR core model, where larger mesh interval sizes precluded examination of spatially-limited heating gradients. These computations provide the basis for thermal-hydraulic analyses to ascertain temperature profiles in the FTR under europia control

  4. Calculation of gamma ray dose buildup factors in water for isotropic point, plane mono directional and line sources using MCNP code

    International Nuclear Information System (INIS)

    Atak, H.; Celikten, O. S.; Tombakoglu, M.

    2009-01-01

    Gamma ray dose buildup factors in water for isotropic point, plane mono directional and infinite/finite line sources were calculated using the MCNP code. The buildup factors are determined for gamma ray energies of 1, 2, 3 and 4 Mev and for shield thicknesses of 1, 2, 4 and 7 mean free paths. The calculated buildup factors were then fitted in the Taylor and Berger forms. For the line sources a buildup factor table was also constructed using the Sievert function and the constants in Taylor form derived in this study to compare with the Monte Carlo results. All buildup factors were compared with the tabulated data given in literature. In order to reduce the statistical errors on buildup factors, 'forced collision' option was used in the MCNP calculations.

  5. Validation of a new midway forward-adjoint coupling option in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I.V.; John, T.M.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-09-01

    The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)

  6. Validation of a new midway forward-adjoint coupling option in MCNP

    International Nuclear Information System (INIS)

    Serov, I.V.; John, T.M.; Hoogenboom, J.E.

    1996-01-01

    The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)

  7. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  8. Fast neutron and gamma-ray transmission technique in mixed samples. MCNP calculations

    International Nuclear Information System (INIS)

    Perez, N.; Padron, I.

    2001-01-01

    In this paper the moisture in sand and also the sulfur content in toluene have been described by using the simultaneous fast neutron/gamma transmission technique (FNGT). Monte Carlo calculations show that it is possible to apply this technique with accelerator-based and isotopic neutron sources in the on-line analysis to perform the product quality control, specifically in the building materials industry and the petroleum one. It has been used particles from a 14MeV neutron generator and also from an Am-Be neutron source. The estimation of optimal system parameters like the efficiency, detection time, hazards and costs were performed in order to compare both neutron sources

  9. Results of calculations of external gamma radiation exposure rates from fallout and the related radionuclide compositions. Operation Tumbler-Snapper, 1952

    International Nuclear Information System (INIS)

    Hicks, H.G.

    1981-07-01

    This report presents data on calculated gamma radiation exposure rates and ground deposition of related radionuclides resulting from Events that deposited detectable radioactivity outside the Nevada Test Site complex

  10. Calculation of the correlation coefficients between the numbers of counts (peak areas and backgrounds) obtained from gamma-ray spectra

    International Nuclear Information System (INIS)

    Korun, M.; Vodenik, B.; Zorko, B.

    2016-01-01

    Two simple methods for calculating the correlations between peaks appearing in gamma-ray spectra are described. We show how the areas are correlated when the peaks do not overlap, but the spectral regions used for the calculation of the background below the peaks do. When the peaks overlap, the correlation can be stronger than in the case of the non-overlapping peaks. The methods presented are simplified to the extent of allowing their implementation with manual calculations. They are intended for practitioners as additional tools to be used when the correlations between the areas of the peaks in the gamma-ray spectra are to be calculated. Also, the correlation coefficient between the number of counts in the peak and the number of counts in the continuous background below the peak is derived. - Highlights: • The correlation coefficients between areas of closely spaced peaks are assessed. • For isolated peaks the correlation arises from the common continuous background. • If peaks overlap the correlation coefficient depends on how much they overlap. • If peaks overlap also the background height affects the correlation coefficient. • The correlation coefficient between the peak area and its background is −1.

  11. ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958

    International Nuclear Information System (INIS)

    Goebelbecker, Hans-Juergen

    2008-01-01

    Description: The papers of the European Atomic Energy Society Symposium VI-58 on radiation shielding (ICRS1) held at Caius College, Cambridge England from 26 to 29 August 1958 are collected here for the first time in electronic form. This symposium was organised in connection with the Second Atoms for Peace Conference held in Geneva Held in Geneva from 1 to 13 September 1958. The Topics discussed covered gamma rays and neutron radiation; the Methods discussed were analytical approaches, semi-empirical Methods, simple computer codes, Monte Carlo method. Little quality nuclear data for shielding calculations was available and the presentations would concentrate on removal cross-sections and build-up factors. Experimental techniques in support to estimate the effective shielding properties of materials were discussed such as general experimental shielding techniques and experiments on neutron attenuation in different materials and on concrete as shield. Foil detectors for spectra measurements and determination of dose rates were mainly used. The typical issues addressed were gamma-heating, gamma spectra, neutron induced gammas, fission products gamma spectra, skyshine radiation and neutron ducts - streaming. Most participants were researchers from the naval and aeronautics sector

  12. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  13. Failure of the Hume-Rothery stabilization mechanism in the Ag5Li8 gamma-brass studied by first-principles FLAPW electronic structure calculations

    International Nuclear Information System (INIS)

    Mizutani, U; Asahi, R; Noritake, T; Sato, H; Takeuchi, T

    2008-01-01

    The first-principles FLAPW (full potential linearized augmented plane wave) electronic structure calculations were performed for the Ag 5 Li 8 gamma-brass, which contains 52 atoms in a unit cell and has been known for many years as one of the most structurally complex alloy phases. The calculations were also made for its neighboring phase AgLi B2 compound. The main objective in the present work is to examine if the Ag 5 Li 8 gamma-brass is stabilized at the particular electrons per atom ratio e/a = 21/13 in the same way as some other gamma-brasses like Cu 5 Zn 8 and Cu 9 Al 4 , obeying the Hume-Rothery electron concentration rule. For this purpose, the e/a value for the Ag 5 Li 8 gamma-brass as well as the AgLi B2 compound was first determined by means of the FLAPW-Fourier method we have developed. It proved that both the gamma-brass and the B2 compound possess an e/a value equal to unity instead of 21/13. Moreover, we could demonstrate why the Hume-Rothery stabilization mechanism fails for the Ag 5 Li 8 gamma-brass and proposed a new stability mechanism, in which the unique gamma-brass structure can effectively lower the band-structure energy by forming heavily populated bonding states near the bottom of the Ag-4d band

  14. Calculational methods for estimating skin dose from electrons in Co-60 gamma-ray beams

    International Nuclear Information System (INIS)

    Higgins, P.D.; Sibata, C.H.; Attix, F.H.; Paliwal, B.R.

    1983-01-01

    Several methods have been employed to calculate the relative contribution to skin dose due to scattered electrons in Co-60 gamma-ray beams. Either the Klein-Nishina differential scattering probability is employed to determine the number and initial energy of electrons scattered into the direction of a detector, or a Gaussian approximation is used to specify the surface distribution of initial pencil electron beams created by parallel or diverging photon fields. Results of these calculations are compared with experimental data. In addition, that fraction of relative surface dose resulting from photon interactions in air alone is estimated and compared with data extrapolated from measurements at large source-surface distance (SSD). The contribution to surface dose from electrons generated in air is 50% or more of the total skin dose for SSDs greater than 80 cm

  15. Experimental verification of methods for gamma dose rate calculations in the vicinity of containers with the RA reactor spent fuel elements

    International Nuclear Information System (INIS)

    Milosevic, M.; Cupac, S.; Pesic, M.

    2005-01-01

    The methodology for equivalent gamma dose rate determination on the outer surface of existing containers with the spent fuel elements of the RA reactor is briefly summarised, and experimental verification of this methodology in the field of gamma rays near the aluminium channel with spent fuel elements lifted from the stainless steel containers no. 275 in the RA reactor hall is presented. The proposed methodology is founded on: the existing fuel burnup data base; methods and models for the photon source determination in the RA reactor spent fuel elements developed in the Vinca Institute, and validated Monte Carlo codes for the equivalent gamma dose rate calculations. (author) [sr

  16. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and {gamma} ray spectrum. FPGS90

    Energy Technology Data Exchange (ETDEWEB)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).

  17. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  18. Double tracer experiments to investigate models for the calculation of gamma doses from a radioactive cloud

    International Nuclear Information System (INIS)

    Nielsen, S.P.; Gryning, S.E.; Thykier-Nielsen, S.; Karlberg, O.; Lyck, E.

    1984-01-01

    The paper presents work from a series of atmospheric dispersion experiments in May 1981 at the Ringhals nuclear power plant in Sweden. The aim of the project was to obtain short-term observations of concentrations and gamma-ray exposures from stack effluents and to compare these results with corresponding values calculated from computer models. Two tracers, sulphurhexafluoride (SF 6 ) and radioactive noble gases, were released from a 110-m stack and detected at ground level downwind at distances of 3-4 km. Calculations were made with two Gaussian plume models: PLUCON developed at Riso National Laboratory and UNIDOSE developed at Studsvik Energiteknik AB. (orig.)

  19. Validation of a model for calculating environmental doses caused by gamma emitters in the soil

    International Nuclear Information System (INIS)

    Ortega, X.; Rosell, J.R.; Dies, X.

    1991-01-01

    A model has been developed to calculate the absorbed dose rates caused by gamma emitters of both natural and artificial origin distributed in the soil. The model divides the soil into five compartments corresponding to layers situated at different depths, and assumes that the concentration of radionuclides is constant in each one of them. The calculations, following the model developed, are undertaken through a program which, based on the concentrations of the radionuclides in the different compartments, gives as a result the dose rate at a height of one metre above the ground caused by each radionuclide and the percentage this represents with respect to the total absorbed dose rate originating from this soil. The validity of the model has been checked in the case of sandy soils by comparing the exposure rates calculated for five sites with the experimental values obtained with an ionisation chamber. (author)

  20. Calculated site substitution in ternary gamma'-Ni3Al: Temperature and composition effects

    DEFF Research Database (Denmark)

    Ruban, Andrei; Skriver, Hans Lomholt

    1997-01-01

    -tin orbitals method in conjunction with the local-density and multisublattice coherent-potential approximations and include all 3d, 4d, 5d, and noble metals. The calculations show the existence of simple trends in the alloying behavior of the gamma' phase which may be explained in a Friedel-like model based...... on the interaction between Ni and the added species. It is shown that the commonly accepted interpretation of the site substitution behavior of Cu and Pd may be incorrect because of site substitution reversal at high temperatures. It is further shown that the direction of the solubility lobe in the ternary phase...

  1. Composite space charge density functions for the calculation of gamma sensitivity of self-powered neutron detectors, using Warren's model

    Science.gov (United States)

    Mahant, A. K.; Rao, P. S.; Misra, S. C.

    1994-07-01

    In the calculational model developed by Warren and Shah for the computation of the gamma sensitivity ( Sγ) it has been observed that the computed Sγ value is quite sensitive to the space charge distribution function assumed for the insulator region and the energy of the gamma photons. The Sγ of SPNDs with Pt, Co and V emitters (manufactured by Thermocoax, France) has been measured at 60Co photon energy and a good correlation between the measured and computed values has been obtained using a composite space charge density function (CSCD), the details of which are presented in this paper. The arguments are extended for evaluating the Sγ values of several SPNDs for which Warren and Shah reported the measured values for a prompt fission gamma spectrum obtained in a swimming pool reactor. These results are also discussed.

  2. Perfecting of shielding calculation technique against the gamma rays arising from a Tokamak with the TFR experience. Application to the conceptual design Tokamak TORE 2 SUPRA

    International Nuclear Information System (INIS)

    Diop, Cheikh M'Backe.

    1980-09-01

    The conception of the necessary shielding around a conceptual design Tokamak requires to execute an estimated calculation of the doses due to the different radiation sources arising from the machine: the thermonuclear neutron source and the gamma ray source emitted during the interaction of the runaway electrons with the diaphragm. In this study, we propose a theorical method to calculate this gamma source. We tackle also the shielding problem of the conceptual design Tokamak: TORE 2 SUPRA [fr

  3. An analytical model to calculate absorbed fractions for internal dosimetry with alpha, beta and gamma emitters

    OpenAIRE

    Amato, Ernesto; Italiano, Antonio; Baldari, Sergio

    2014-01-01

    We developed a general model for the calculation of absorbed fractions in ellipsoidal volumes of soft tissue uniformly filled with alpha, beta and gamma emitting radionuclides. The approach exploited Monte Carlo simulations with the Geant4 code to determine absorbed fractions in ellipsoids characterized by a wide range of dimensions and ellipticities, for monoenergetic emissions of each radiation type. The so-obtained absorbed fractions were put in an analytical relationship with the 'general...

  4. Determination of gamma production from (n, gamma) reactions

    International Nuclear Information System (INIS)

    Kostal, M.

    2007-06-01

    Calculation of gamma production by interaction of neutrons with materials requires a reasonable accuracy of the nuclear libraries, i. e. effective cross sections, nuclear levels and probabilities of transitions between them. Accurate data enable accurate calculations to be performed, e.g. for PGNAA. First, gamma production in a thick 56 Fe target was examined. Appreciable discrepancies were found among the nuclear libraries available. Additional calculations were performed and compared with the observed data. The fluence of photons observed behind a thick iron target was investigated, the target being irradiated with neutrons from the front side. The results were evaluated for the various nuclear libraries. It is concluded that the libraries ENDF/B VI.2., i.e. data embedded in the MCNPX code, are sufficient for a number of applications. However, their accuracy is insufficient for prompt gamma neutron activation analysis. This is also true of data from the libraries JEFF 3.1. a JENDL 3.3, so that other libraries will have to be used for PGNAA. Specifically for 56 Fe, the data from the libraries ENDF/B VII.0 seem to be usable. (P.A.)

  5. Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

    Science.gov (United States)

    Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

    2018-03-01

    In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

  6. EGSnrc calculated and MRI-polymer gel dosimeter measured dose distribution of gamma knife in presence of inhomogeneities

    International Nuclear Information System (INIS)

    Allahverdi Pourfallah, T.; Allahverdi, M.; Riahi Alam, N.; Ay, M.; Zahmatkesh, M.; Ibbott, J.S.

    2008-01-01

    Stereotactic gamma-knife radiosurgery plays an important role in managing small intracranial brain lesions. Currently, polymer gel dosimetry is still the only dosimetry method for directly measuring three-dimensional dose distributions. polymer gel dosimeters are tissue equivalent and can act as a phantom material. In this study effects of inhomogeneities on those distributions have been investigated using both EGSnrc calculation and PAGAT polymer gel dosimeter. (author)

  7. GammaModeler 3-D gamma-ray imaging technology

    International Nuclear Information System (INIS)

    2000-01-01

    The 3-D GammaModelertrademark system was used to survey a portion of the facility and provide 3-D visual and radiation representation of contaminated equipment located within the facility. The 3-D GammaModelertrademark system software was used to deconvolve extended sources into a series of point sources, locate the positions of these sources in space and calculate the 30 cm. dose rates for each of these sources. Localization of the sources in three dimensions provides information on source locations interior to the visual objects and provides a better estimate of the source intensities. The three dimensional representation of the objects can be made transparent in order to visualize sources located within the objects. Positional knowledge of all the sources can be used to calculate a map of the radiation in the canyon. The use of 3-D visual and gamma ray information supports improved planning decision-making, and aids in communications with regulators and stakeholders

  8. Modeling the irradiation facility in the Deir Al-Hajar area to calculate the spatial gamma dose distribution using the MCNP code

    International Nuclear Information System (INIS)

    Khattab, K.; Bush, M; Kassery, H.

    2009-03-01

    A 3-D model for the irradiation plant which belongs to the Atomic Energy Commission, Department of Radiation Technology in the Deir Al-Hajar area near Damascus, is presented in this work using the MCNP-4C code. This model is used to calculate the spatial gamma ray dose in the (x, y, z) coordinate. Good agreements are noticed between the measured and the calculated results. (author)

  9. Computer model for calculating gamma-ray pulse-height spectra for logging applications

    International Nuclear Information System (INIS)

    Evans, M.L.

    1981-01-01

    A generalized computer model has been devised to simulate the emission, transport, and detection of natural gamma radiation from various logging environments. The model yields high-resolution gamma-ray pulse-height spectra that can be used to correct both gross gamma and spectral gamma-ray logs. The technique can help provide corrections to airborne and surface radiometric survey logs for the effects of varying altitude, formation composition, and overburden. Applied to borehole logging, the model can yield estimates of the effects of varying borehole fluid and casing attenuations, as well as varying formation porosity and saturation

  10. Methods of calculation of cross section of reaction 115In(gamma, n)114mIn

    International Nuclear Information System (INIS)

    Zhaba, V.I.; Parlag, A.M.

    2015-01-01

    The cross section of reaction 115 In(gamma, n) 114m In is expected by different methods. Results of the got cross section it is well comported inter se the Penfold-Leiss and Tikhonov's methods. The calculation of cross section is conducted the Penfold-Leiss method with smoothing out by the method of iterations. Number of iterations n = 1; 3; 5. In the programmatic package of TALYS-1.4 got cross section for five models of closeness of levels. Theoretical and experimental results well coincide in a maximum.

  11. Use of gamma spectroscopy in activation analysis; Utilisation de la spectrographie gamma dans l'analyse par activation

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    Brief review of the principles of activation analysis: calculation of activities, decay curves, {beta} absorption curves, examples of application. - Principle and description of the {gamma} spectrograph. - Practical utilisation of the {gamma} spectrograph: analysis by activation, analysis by {beta} - x fluorescence. - Sensitivity limit of the method and precision of the measurements. - Possible improvements to the method: {gamma} spectroscopy with elimination of the Compton effect. (author) [French] Bref rappel des principes de l'analyse par activation: calcul des activites, courbes de decroissance, courbes d'absorption {beta}, exemples d'utilisation. - Principe et description du spectrographe {gamma}. - Utilisation pratique de la spectrographie {gamma}: analyse par activation, analyse par fluorescence {beta} - x. - Limite de sensibilite de la methode et precision des mesures. - Ameliorations possibles de la methode: spectrographe {gamma} avec elimination de l'effet Compton. (auteur)

  12. BFKL resummation effects in gamma* gamma* to rho rho

    Energy Technology Data Exchange (ETDEWEB)

    Enberg, R.; Pire, B.; Szymanowski, L.; Wallon, S.

    2005-08-11

    We calculate the leading order BFKL amplitude for the exclusive diffractive process {gamma}*{sub L}(Q{sub 1}{sup 2}) {gamma}*{sub L}(Q{sub 2}{sup 2}) {yields} {rho}{sub L}{sup 0}{rho}{sub L}{sup 0} in the forward direction, which can be studied in future high energy e{sup +}e{sup -} linear colliders. The resummation effects are very large compared to the fixed-order calculation. We also estimate the next-to-leading logarithmic corrections to the amplitude by using a specific resummation of higher order effects and find a substantial growth with energy, but smaller than in the leading logarithmic approximation.

  13. Application of the decoupling scheme on complex neutron-gamma shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

    2000-03-01

    Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

  14. Measurement of the Charm Production Cross Section in gamma gamma Collisions at LEP

    CERN Document Server

    Acciarri, M.; Adriani, O.; Aguilar-Benitez, M.; Alcaraz, J.; Alemanni, G.; Allaby, J.; Aloisio, A.; Alviggi, M.G.; Ambrosi, G.; Anderhub, H.; Andreev, Valery P.; Angelescu, T.; Anselmo, F.; Arefev, A.; Azemoon, T.; Aziz, T.; Bagnaia, P.; Bajo, A.; Baksay, L.; Balandras, A.; Baldew, S.V.; Banerjee, S.; Banerjee, Sw.; Barczyk, A.; Barillere, R.; Bartalini, P.; Basile, M.; Batalova, N.; Battiston, R.; Bay, A.; Becattini, F.; Becker, U.; Behner, F.; Bellucci, L.; Berbeco, R.; Berdugo, J.; Berges, P.; Bertucci, B.; Betev, B.L.; Bhattacharya, S.; Biasini, M.; Biland, A.; Blaising, J.J.; Blyth, S.C.; Bobbink, G.J.; Bohm, A.; Boldizsar, L.; Borgia, B.; Bourilkov, D.; Bourquin, M.; Braccini, S.; Branson, J.G.; Brochu, F.; Buffini, A.; Buijs, A.; Burger, J.D.; Burger, W.J.; Cai, X.D.; Capell, M.; Cara Romeo, G.; Carlino, G.; Cartacci, A.M.; Casaus, J.; Castellini, G.; Cavallari, F.; Cavallo, N.; Cecchi, C.; Cerrada, M.; Cesaroni, F.; Chamizo, M.; Chang, Y.H.; Chaturvedi, U.K.; Chemarin, M.; Chen, A.; Chen, G.; Chen, G.M.; Chen, H.F.; Chen, H.S.; Chiefari, G.; Cifarelli, L.; Cindolo, F.; Civinini, C.; Clare, I.; Clare, R.; Coignet, G.; Colino, N.; Costantini, S.; Cotorobai, F.; de la Cruz, B.; Csilling, A.; Cucciarelli, S.; Dai, T.S.; van Dalen, J.A.; D'Alessandro, R.; de Asmundis, R.; Deglon, P.; Degre, A.; Deiters, K.; della Volpe, D.; Delmeire, E.; Denes, P.; DeNotaristefani, F.; De Salvo, A.; Diemoz, M.; Dierckxsens, M.; van Dierendonck, D.; Dionisi, C.; Dittmar, M.; Dominguez, A.; Doria, A.; Dova, M.T.; Duchesneau, D.; Dufournaud, D.; Duinker, P.; El Mamouni, H.; Engler, A.; Eppling, F.J.; Erne, F.C.; Ewers, A.; Extermann, P.; Fabre, M.; Falagan, M.A.; Falciano, S.; Favara, A.; Fay, J.; Fedin, O.; Felcini, M.; Ferguson, T.; Fesefeldt, H.; Fiandrini, E.; Field, J.H.; Filthaut, F.; Fisher, P.H.; Fisk, I.; Forconi, G.; Freudenreich, K.; Furetta, C.; Galaktionov, Iouri; Ganguli, S.N.; Garcia-Abia, Pablo; Gataullin, M.; Gau, S.S.; Gentile, S.; Gheordanescu, N.; Giagu, S.; Gong, Z.F.; Grenier, Gerald Jean; Grimm, O.; Gruenewald, M.W.; Guida, M.; van Gulik, R.; Gupta, V.K.; Gurtu, A.; Gutay, L.J.; Haas, D.; Hasan, A.; Hatzifotiadou, D.; Hebbeker, T.; Herve, Alain; Hidas, P.; Hirschfelder, J.; Hofer, H.; Holzner, G.; Hoorani, H.; Hou, S.R.; Hu, Y.; Iashvili, I.; Jin, B.N.; Jones, Lawrence W.; de Jong, P.; Josa-Mutuberria, I.; Khan, R.A.; Kafer, D.; Kaur, M.; Kienzle-Focacci, M.N.; Kim, D.; Kim, J.K.; Kirkby, Jasper; Kiss, D.; Kittel, W.; Klimentov, A.; Konig, A.C.; Kopal, M.; Kopp, A.; Koutsenko, V.; Kraber, M.; Kraemer, R.W.; Krenz, W.; Kruger, A.; Kunin, A.; Ladron de Guevara, P.; Laktineh, I.; Landi, G.; Lebeau, M.; Lebedev, A.; Lebrun, P.; Lecomte, P.; Lecoq, P.; Le Coultre, P.; Lee, H.J.; Le Goff, J.M.; Leiste, R.; Levtchenko, P.; Li, C.; Likhoded, S.; Lin, C.H.; Lin, W.T.; Linde, F.L.; Lista, L.; Liu, Z.A.; Lohmann, W.; Longo, E.; Lu, Y.S.; Lubelsmeyer, K.; Luci, C.; Luckey, David; Lugnier, L.; Luminari, L.; Lustermann, W.; Ma, W.G.; Maity, M.; Malgeri, L.; Malinin, A.; Mana, C.; Mangeol, D.; Mans, J.; Marian, G.; Martin, J.P.; Marzano, F.; Mazumdar, K.; McNeil, R.R.; Mele, S.; Merola, L.; Meschini, M.; Metzger, W.J.; von der Mey, M.; Mihul, A.; Milcent, H.; Mirabelli, G.; Mnich, J.; Mohanty, G.B.; Moulik, T.; Muanza, G.S.; Muijs, A.J.M.; Musicar, B.; Musy, M.; Napolitano, M.; Nessi-Tedaldi, F.; Newman, H.; Niessen, T.; Nisati, A.; Kluge, Hannelies; Ofierzynski, R.; Organtini, G.; Oulianov, A.; Palomares, C.; Pandoulas, D.; Paoletti, S.; Paolucci, P.; Paramatti, R.; Park, H.K.; Park, I.H.; Passaleva, G.; Patricelli, S.; Paul, Thomas Cantzon; Pauluzzi, M.; Paus, C.; Pauss, F.; Pedace, M.; Pensotti, S.; Perret-Gallix, D.; Petersen, B.; Piccolo, D.; Pierella, F.; Pieri, M.; Piroue, P.A.; Pistolesi, E.; Plyaskin, V.; Pohl, M.; Pojidaev, V.; Postema, H.; Pothier, J.; Prokofev, D.O.; Prokofiev, D.; Quartieri, J.; Rahal-Callot, G.; Rahaman, M.A.; Raics, P.; Raja, N.; Ramelli, R.; Rancoita, P.G.; Ranieri, R.; Raspereza, A.; Raven, G.; Razis, P.; Ren, D.; Rescigno, M.; Reucroft, S.; Riemann, S.; Riles, Keith; Rodin, J.; Roe, B.P.; Romero, L.; Rosca, A.; Rosier-Lees, S.; Roth, Stefan; Rosenbleck, C.; Roux, B.; Rubio, J.A.; Ruggiero, G.; Rykaczewski, H.; Saremi, S.; Sarkar, S.; Salicio, J.; Sanchez, E.; Sanders, M.P.; Schafer, C.; Schegelsky, V.; Schmidt-Kaerst, S.; Schmitz, D.; Schopper, H.; Schotanus, D.J.; Schwering, G.; Sciacca, C.; Seganti, A.; Servoli, L.; Shevchenko, S.; Shivarov, N.; Shoutko, V.; Shumilov, E.; Shvorob, A.; Siedenburg, T.; Son, D.; Smith, B.; Spillantini, P.; Steuer, M.; Stickland, D.P.; Stone, A.; Stoyanov, B.; Straessner, A.; Sudhakar, K.; Sultanov, G.; Sun, L.Z.; Sushkov, S.; Suter, H.; Swain, J.D.; Szillasi, Z.; Sztaricskai, T.; Tang, X.W.; Tauscher, L.; Taylor, L.; Tellili, B.; Teyssier, D.; Timmermans, Charles; Ting, Samuel C.C.; Ting, S.M.; Tonwar, S.C.; Toth, J.; Tully, C.; Tung, K.L.; Uchida, Y.; Ulbricht, J.; Valente, E.; Vesztergombi, G.; Vetlitsky, I.; Vicinanza, D.; Viertel, G.; Villa, S.; Vivargent, M.; Vlachos, S.; Vodopianov, I.; Vogel, H.; Vogt, H.; Vorobev, I.; Vorobov, A.A.; Vorvolakos, A.; Wadhwa, M.; Wallraff, W.; Wang, M.; Wang, X.L.; Wang, Z.M.; Weber, A.; Weber, M.; Wienemann, P.; Wilkens, H.; Wu, S.X.; Wynhoff, S.; Xia, L.; Xu, Z.Z.; Yamamoto, J.; Yang, B.Z.; Yang, C.G.; Yang, H.J.; Yang, M.; Ye, J.B.; Yeh, S.C.; Zalite, A.; Zalite, Yu.; Zhang, Z.P.; Zhu, G.Y.; Zhu, R.Y.; Zichichi, A.; Zilizi, G.; Zimmermann, B.; Zoller, M.

    2001-01-01

    Open charm production in gamma-gamma collisions is studied with data collected at e+e- centre-of-mass energies from 189 GeV to 202 GeV corresponding to a total integrated luminosity of 410 pb-1. The charm cross section sigma(gamma gamma ---> c c~ X) is measured for the first time as a function of the two-photon centre-of-mass energy in the interval from 5 GeV to 70 GeV and is compared to NLO QCD calculations.

  15. Reaction /sup 56/Fe (. gamma. ,. cap alpha. /sub 0/) and /sup 56/Fe (. gamma. , p/sub 0/)

    Energy Technology Data Exchange (ETDEWEB)

    Tamae, T; Sugawara, M [Tohoku Univ., Sendai (Japan). Lab. of Nuclear Science; Tsubota, H

    1975-06-01

    Precise analysis was made on the cross section of the /sup 56/Fe (..gamma.., ..cap alpha../sub 0/) reaction and the angular distribution at Esub(e) = 17 MeV, including the systematic error. The (..gamma.., ..cap alpha../sub 0/) reaction cross section was compared with a calculation using the compound nucleus model, utilizing the photon absorption cross section derived from the experimental values of /sup 56/Fe (..gamma.., n) and /sup 56/Fe (..gamma.., p) cross sections. From the (..gamma.., ..cap alpha../sub 0/) reaction cross section data of various nuclei, an empirical formula was obtained for determining the position of a peak in the (..gamma.., ..cap alpha../sub 0/) reaction cross section. The /sup 56/Fe (..gamma.., p/sub 0/) reaction cross section measured at an excitation energy in the range of 14.6--25.0 MeV was compared with the calculated one with the compound nucleus model, but the form and size differ totally.

  16. Dosimetric calculations by Monte Carlo for treatments of radiosurgery with the Leksell Gamma Knife, homogeneous and non homogeneous cases; Calculos dosimetricos por Monte Carlo para tratamientos de radiocirugia con el Leksell Gamma Knife, casos homogeneo y no homogeneo

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Lallena R, A.M. [Universidad de Granada (Spain)

    2004-07-01

    In this work dose profiles are calculated that are obtained modeling treatments of radiosurgery with the Leksell Gamma Knife. This was made with the simulation code Monte Carlo Penelope for an homogeneous mannequin and one not homogeneous. Its were carried out calculations with the irradiation focus coinciding with the center of the mannequin as in near areas to the bone interface. Each one of the calculations one carries out for the 4 skull treatment that it includes the Gamma Knife and using a model simplified of their 201 sources of {sup 60} Co. It was found that the dose profiles differ of the order of 2% when the isocenter coincides with the center of the mannequin and they ascend to near 5% when the isocenter moves toward the skull. (Author)

  17. Methodology of shielding calculation for nuclear reactors

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

    1982-01-01

    A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

  18. Gamma-ray streaming in bent ducts and voids

    International Nuclear Information System (INIS)

    Bourdet, L.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    We have developed an analytical method to calculate gamma-ray streaming through straight ducts and a numerical method to study the gamma propagation in bends or in annular clearances. The whole set allows a rigorous treatment of gamma streaming through bent ducts. In the same time a Monte Carlo method allows to study any form of geometry, by using sophisticated biasing techniques. All these developments are made with a simplified albedo. An easy to use code is also proposed to calculate very general albedos and a code to calculate the dose rate due to reflection in a room. Gamma dose rate albedos are determined for all elements and the energy range which concerns fission reactors

  19. Poker-camp: a program for calculating detector responses and phantom organ doses in environmental gamma fields

    International Nuclear Information System (INIS)

    Koblinger, L.

    1981-09-01

    A general description, user's manual and a sample problem are given in this report on the POKER-CAMP adjoint Monte Carlo photon transport program. Gamma fields of different environmental sources which are uniformly or exponentially distributed sources or plane sources in the air, in the soil or in an intermediate layer placed between them are simulated in the code. Calculations can be made on flux, kerma and spectra of photons at any point; and on responses of point-like, cylindrical, or spherical detectors; and on doses absorbed in anthropomorphic phantoms. (author)

  20. Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

    Science.gov (United States)

    Brovchenko, Mariya; Duhamel, Isabelle; Dechenaux, Benjamin

    2017-09-01

    The present paper presents the study carried out in the frame of the DISCOMS project, which stands for "DIstributed Sensing for COrium Monitoring and Safety". This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

  1. Parameters calculation of a shielding experiment and evaluation of calculation methodology

    International Nuclear Information System (INIS)

    Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.

    1986-01-01

    In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt

  2. Nuclear models and data for gamma-ray production

    International Nuclear Information System (INIS)

    Young, P.G.

    1975-01-01

    The current Evaluated Nuclear Data File (ENDF/B, Version IV) contains information on prompt gamma-ray production from neutron-induced reactions for some 38 nuclides. In addition, there is a mass of fission product yield, capture, and radioactive decay data from which certain time-dependent gamma-ray results can be calculated. These data are needed in such applications as gamma-ray heating calculations for reactors, estimates of radiation levels near nuclear facilities and weapons, shielding design calculations, and materials damage estimates. The prompt results are comprised of production cross sections, multiplicities, angular distributions, and energy spectra for secondary gamma-rays from a variety of reactions up to an incident neutron energy of 20 MeV. These data are based in many instances on experimental measurements, but nuclear model calculations, generally of a statistical nature, are also frequently used to smooth data, to interpolate between measurements, and to calculate data in unmeasured regions. The techniques and data used in determining the ENDF/B evaluations are reviewed, and comparisons of model-code calculations and ENDF data with recent experimental results are given. 11 figures

  3. A versatile program for the calculation of linear accelerator room shielding.

    Science.gov (United States)

    Hassan, Zeinab El-Taher; Farag, Nehad M; Elshemey, Wael M

    2018-03-22

    This work aims at designing a computer program to calculate the necessary amount of shielding for a given or proposed linear accelerator room design in radiotherapy. The program (Shield Calculation in Radiotherapy, SCR) has been developed using Microsoft Visual Basic. It applies the treatment room shielding calculations of NCRP report no. 151 to calculate proper shielding thicknesses for a given linear accelerator treatment room design. The program is composed of six main user-friendly interfaces. The first enables the user to upload their choice of treatment room design and to measure the distances required for shielding calculations. The second interface enables the user to calculate the primary barrier thickness in case of three-dimensional conventional radiotherapy (3D-CRT), intensity modulated radiotherapy (IMRT) and total body irradiation (TBI). The third interface calculates the required secondary barrier thickness due to both scattered and leakage radiation. The fourth and fifth interfaces provide a means to calculate the photon dose equivalent for low and high energy radiation, respectively, in door and maze areas. The sixth interface enables the user to calculate the skyshine radiation for photons and neutrons. The SCR program has been successfully validated, precisely reproducing all of the calculated examples presented in NCRP report no. 151 in a simple and fast manner. Moreover, it easily performed the same calculations for a test design that was also calculated manually, and produced the same results. The program includes a new and important feature that is the ability to calculate required treatment room thickness in case of IMRT and TBI. It is characterised by simplicity, precision, data saving, printing and retrieval, in addition to providing a means for uploading and testing any proposed treatment room shielding design. The SCR program provides comprehensive, simple, fast and accurate room shielding calculations in radiotherapy.

  4. Integral transport computation of gamma detector response with the CPM2 code

    International Nuclear Information System (INIS)

    Jones, D.B.

    1989-12-01

    CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

  5. Air-over-ground calculations of the neutron, prompt, and secondary-gamma free-in-air tissue kerma from the Hiroshima and Nagasaki devices

    International Nuclear Information System (INIS)

    Pace, J.V. III; Knight, J.R.; Bartine, D.E.

    1982-01-01

    This paper reports preliminary results of the two-dimensional discrete-ordinate, calculations for the air-over-ground transport of radiation from the Hiroshima and Nagasaki weapon devices. It was found that the gamma-ray kerma dominated the total kerma for both environments

  6. KaKs_Calculator 2.0: A Toolkit Incorporating Gamma-Series Methods and Sliding Window Strategies

    KAUST Repository

    Wang, Dapeng

    2010-05-05

    We present an integrated stand-alone software package named KaKs_Calculator 2.0 as an updated version. It incorporates 17 methods for the calculation of nonsynonymous and synonymous substitution rates; among them, we added our modified versions of several widely used methods as the gamma series including γ-NG, γ-LWL, γ-MLWL, γ-LPB, γ-MLPB, γ-YN and γ-MYN, which have been demonstrated to perform better under certain conditions than their original forms and are not implemented in the previous version. The package is readily used for the identification of positively selected sites based on a sliding window across the sequences of interests in 5\\' to 3\\' direction of protein-coding sequences, and have improved the overall performance on sequence analysis for evolution studies. A toolbox, including C++ and Java source code and executable files on both Windows and Linux platforms together with a user instruction, is downloadable from the website for academic purpose at https://sourceforge.net/projects/kakscalculator2/.

  7. KaKs_Calculator 2.0: A Toolkit Incorporating Gamma-Series Methods and Sliding Window Strategies

    KAUST Repository

    Wang, Dapeng; Zhang, Yubin; Zhang, Zhang; Zhu, Jiang; Yu, Jun

    2010-01-01

    We present an integrated stand-alone software package named KaKs_Calculator 2.0 as an updated version. It incorporates 17 methods for the calculation of nonsynonymous and synonymous substitution rates; among them, we added our modified versions of several widely used methods as the gamma series including γ-NG, γ-LWL, γ-MLWL, γ-LPB, γ-MLPB, γ-YN and γ-MYN, which have been demonstrated to perform better under certain conditions than their original forms and are not implemented in the previous version. The package is readily used for the identification of positively selected sites based on a sliding window across the sequences of interests in 5' to 3' direction of protein-coding sequences, and have improved the overall performance on sequence analysis for evolution studies. A toolbox, including C++ and Java source code and executable files on both Windows and Linux platforms together with a user instruction, is downloadable from the website for academic purpose at https://sourceforge.net/projects/kakscalculator2/.

  8. In-situ gamma spectrometry method for determination of environmental gamma dose

    International Nuclear Information System (INIS)

    Conti, Claudio de Carvalho

    1995-07-01

    This work tries to establish a methodology for germanium detectors calibration, normally used for in situ gamma ray spectrometry, for determining the environmental exposure rate in function of the energy of the incident photons. For this purpose a computer code has been developed, based on the stripping method, for the computational spectra analysis to calculate the contribution of the partial absorption of the gamma rays (Compton effect) in the active and nonactive parts of the detector. The resulting total absorption spectrum is then converted to fluence distribution in function of the energy for the photons reaching the detector, which is then used to calculate the exposure rate or kerma in air. The unfolding and fluency convention parameters are determined by detector calibration using point gamma sources. The method is validated by comparison of the results against the calculated exposure rate at a point of interest for the standards. This method is used for the direct measurement of the exposure rate distribution in function of the energy at the site, in situ measurement technic, leading to rapid results during an emergency situation and also used for indoor measurements. (author)

  9. Formulation of the relationship between indices of neutron-gamma and gamma-gamma method and the percentrage of iron

    International Nuclear Information System (INIS)

    Majorowicz, J.

    1973-01-01

    In this article, the author presents the possibility of a complex utilization of radiometric logging methods, neutron-gamma profiling and gamma-gamma density logging for determining percentage of iron and establishing geophysical possibilities of identifying zones of economically profitable ores in borehole profiles. Figures present the correlations between indices of neutron-gamma and gamma-gamma logging methods and the percentage of iron, as well as the correlation of neutron-gamma and gamma-gamma indices for zones minerallized with iron ores. The article presents the correlational analyses of the results: the correlational coefficients are given as well as total error in determining iron content on the basis of each of the methods described. Next, a multidimensional statistical analysis is carried out on the results obtained. On the basis of the two-dimensional correlational coefficients calculated and the average standard deviation, an equation of linear regression was formulated, simultaneously involving three parameters - the indices of neutron-gamma and gamma-gamma logging and the percentage of iron. The multiple correlational coefficient obtained markedly exceeds the two-dimentional correlation coefficient (r=0.974>rsub(xz)>rsub(yz)>rsub(xy)). The given method of utilizing multidimensional statistics in borehole geophysics for identifying iron ores is an efficient one. On the basis of several relationships among independent variables which are less obvious (smaller values of correlational coefficient), it is possible to obtain a single distinct relationship involving all variables simultaneously. (author)

  10. Nondestructive analysis of the RA fuel burnup, Calculation of the gamma activity ratio of fission products in the fuel - program QU0C1

    International Nuclear Information System (INIS)

    Bulovic, V.F.

    1973-01-01

    The γ radiation of RA reactor fuel element was measured under precisely defined measuring conditions. The spectrum was analysed by spectrometer with semiconductor Ge(Li) detector. The gamma counting rate in the fuel spectrum is defined as a function of fission product activity, gamma energy and yield, fuel thickness and additional absorbers, dimensions of the gamma collimator. Activity ratio of two fission products is defined as a function of counting rate peaks and part of the mentioned quantities. Four options for calculating the activities for fission products are discussed. Three of them are covered by the QU0C1 code written in FORTRAN for the CDC 3600 computer. The code is included in this report [sr

  11. BFKL resummation effects in {gamma}{sup *}{gamma}{sup *}{yields}{rho}{rho}

    Energy Technology Data Exchange (ETDEWEB)

    Enberg, R. [Ecole Polytechnique, CPHT, Palaiseau (France); Lawrence Berkeley National Laboratory, Berkeley (United States); Pire, B. [Ecole Polytechnique, CPHT, Palaiseau (France); Szymanowski, L. [Soltan Institute for Nuclear Studies, Warsaw (Poland); Universite de Liege, Liege (Belgium); Wallon, S. [LPT, Universite Paris-Sud, Orsay (France)

    2006-03-15

    We calculate the leading order BFKL amplitude for the exclusive diffractive process {gamma}{sup *}{sub L}(Q{sub 1}{sup 2}){gamma}{sup *}{sub L}(Q{sub 2}{sup 2}){yields}{rho}{sub L}{sup 0}{rho}{sub L}{sup 0} in the forward direction, which can be studied in future high energy e{sup +}e{sup -} linear colliders. The resummation effects are very large compared to the fixed-order calculation. We also estimate the next-to-leading logarithmic corrections to the amplitude by using a specific resummation of higher order effects and find a substantial growth with energy, but smaller than in the leading logarithmic approximation. (orig.)

  12. Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

    International Nuclear Information System (INIS)

    Anspaugh, L.R.; Church, B.W.

    1986-01-01

    In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, UT; Ely, NV; and Las Vegas, NV. Three events, HARRY (19 May 1953), BEE (22 March 1955), and SMOKY (31 August 1957), accounted for more than half the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of infinite exposure, estimated exposure, and 1-yr effective biological exposure are explained

  13. Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

    International Nuclear Information System (INIS)

    Anspaugh, L.R.; Church, B.W.

    1985-12-01

    In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, Utah; Ely, Nevada; and Las Vegas, Nevada. Three events, HARRY (May 19, 1953), BEE (March 22, 1955), and SMOKY (August 31, 1957), accounted for over half of the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of ''infinite exposure,'' ''estimated exposure,'' and ''one year effective biological exposure'' are explained. 4 figs., 7 tabs

  14. An assessment of the feasibility of using Monte Carlo calculations to model a combined neutron/gamma electronic personal dosemeter

    International Nuclear Information System (INIS)

    Tanner, J.E.; Witts, D.; Tanner, R.J.; Bartlett, D.T.; Burgess, P.H.; Edwards, A.A.; More, B.R.

    1995-01-01

    A Monte Carlo facility has been developed for modelling the response of semiconductor devices to mixed neutron-photon fields. This utilises the code MCNP for neutron and photon transport and a new code, STRUGGLE, which has been developed to model the secondary charged particle transport. It is thus possible to predict the pulse height distribution expected from prototype electronic personal detectors, given the detector efficiency factor. Initial calculations have been performed on a simple passivated implanted planar silicon detector. This device has also been irradiated in neutron, gamma and X ray fields to verify the accuracy of the predictions. Good agreement was found between experiment and calculation. (author)

  15. Deposited power in a complex device by gamma radiation of test reactors; experiments and calculations carried out at SILOE

    International Nuclear Information System (INIS)

    Petitcolas, H.; Besson, A.; Bevilacqua, A.; Cosoli, G.

    1984-09-01

    Eight samples, which represent different materials used in testing reactors, were irradiated in the device ''CYRANO'' placed in the water reflector at different distances from the reactor core. The power dissipated in the device was measured by the ''CYRANO'' equipment itself, whereas the calorimeter juxtaposed served to monitor the gamma flux. Parallel to each experiment, the power deposited in the samples, the device materials and the calorimeter was calculated by the code MERCURE 4. The measured values were compared with the calculated ones, both in relative and in absolute values, for each sample and for each distance in the reflector. The comparison shows very good agreement [fr

  16. Gamma-ray production cross sections for MeV neutrons

    International Nuclear Information System (INIS)

    Kitazawa, Hideo; Harima, Yoshiko; Yamakoshi, Hisao; Sano, Yuji; Kobayashi, Tsuguyuki.

    1979-01-01

    Gamma-ray production cross section and spectra for 1- to 20-MeV neutrons were theoretically obtained, which were requested for heating calculations, for shielding design calculations, and for material damage estimates. Calculations were carried out for Al, Si, Ca, Fe, Ni, Cu, Nb, Ta, Au, and Pb, using a spin-dependent evaporation model without the parity conservation and including the dipole and quardupole gamma-ray transitions. The results were compared with the experimental data measured in ORNL to confirm the availability of this model in applications. In addition, the effects on the gamma-ray production cross section of the optical potential, level density, yrast level, and radiation width were investigated in detail. The conclusions are: 1) the use of the optical potential which gives the correct total reaction cross section is essential to gamma-ray production calculations, 2) the gamma-ray production cross section is not so sensitive to the choice of level density parameters, 3) the inclusion of yrast levels is necessary in dealing with the competition of the neutron and gamma-ray emissions from highly excited states, and 4) the Brink-Axel type's radiation width is unsuitable to be applied to radiative capture processes. (author)

  17. ECP measurements under neutron and gamma ray in in-pile loop and their data evaluation by water radiolysis calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic)

    2010-07-01

    In order to establish reliable electrochemical corrosion potential (ECP) sensors for applying in reactor core peripherals of power plants, performance tests of sensors under irradiation were carried out in the in-pile loop of the experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Responses of different kinds of sensors under neutron and gamma irradiation conditions have been compared each other. Corrosive conditions along the in-pile loop were calculated by water radiolysis calculation code, WRAC-J and calculated corrosive conditions were compared with the measured results. As a result of the evaluation, it was confirmed that the ECP sensors could be applied to irradiation conditions of reactor peripherals, while the water radiolysis model could be also applied for evaluation of corrosive conditions of reactor peripherals. (author)

  18. Calculation of critical level value for radioactivity detection in gamma spectrometric analysis on the base of semiconductor detectors under the Chernobyl' conditions in 1986-1987

    International Nuclear Information System (INIS)

    Glazunov, V.O.; Rusyaev, R.V.

    1989-01-01

    The problem of determination of radioactivity critical level in a sample by means of gamma spectrometer with semiconductor detector is studied theoretically. The formula for critical level, which shows that it is necessary to know the background pulse counting rate in order to determine the minimum gamma photon pulse counting rates, is derived. Calculations of critical level for the Chernobyl' conditions in time period from October 1986 till July 1987 are made. 8 refs.; 7 figs.; 17 tabs

  19. LOFT gamma densitometer background fluxes

    International Nuclear Information System (INIS)

    Grimesey, R.A.; McCracken, R.T.

    1978-01-01

    Background gamma-ray fluxes were calculated at the location of the γ densitometers without integral shielding at both the hot-leg and cold-leg primary piping locations. The principal sources for background radiation at the γ densitometers are 16 N activity from the primary piping H 2 O and γ radiation from reactor internal sources. The background radiation was calculated by the point-kernel codes QAD-BSA and QAD-P5A. Reasonable assumptions were required to convert the response functions calculated by point-kernel procedures into the gamma-ray spectrum from reactor internal sources. A brief summary of point-kernel equations and theory is included

  20. Total absorption gamma-ray spectroscopy (TAGS): Current status of measurement programmes for decay heat calculations and other applications. Summary report of consultants' meeting

    International Nuclear Information System (INIS)

    Nichols, A.L.; Nordborg, C.

    2009-02-01

    A Consultants' Meeting on 'Total Absorption Gamma-ray Spectroscopy (TAGS)' was held on 27-28 January 2009 at the IAEA Headquarters, Vienna, Austria. All presentations, discussions and recommendations of this meeting are contained within this report. The purpose of the meeting was to report and discuss progress and plans to measure total gamma-ray spectra in order to derive mean beta and gamma decay data for decay heat calculations and other applications. This form of review had been recommended by contributors to Subgroup 25 of the OECD-NEA Working Party on International Evaluation Cooperation of the Nuclear Science Committee, for implementation in 2008/09. Hence, relevant specialists were invited to discuss their recently performed and planned TAGS studies, along with experimentalists proposing to assemble and operate such dedicated facilities. Knowledge and quantification of antineutrino spectra is believed to be a significant asset in the non-invasive monitoring of reactor operations and possible application in safeguards, as well as fundamental in the study of neutrino oscillations - these data needs were also debated in terms of appropriate TAGS measurements. A re-assessment of the current request list for TAGS studies is merited and was undertaken in the context of decay heat calculations, and agreement was reached to extend these requirements to the derivation of antineutrino spectra. (author)

  1. Analysis of ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}D{sub 2}){gamma}{sub 5g}, {gamma}{sub 3g} and ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}L{sub 6}){gamma}{sub 1g}, a{gamma}{sub 5g} two-photon absorption spectra of Cs{sub 2}NaYF{sub 6}:Eu{sup 3+}

    Energy Technology Data Exchange (ETDEWEB)

    Ning Lixin; Wang Dianyuan; Xia Shangda [Structure Research Laboratory, Academica Sinica, Department of Physics, University of Science and Technology of China, Heifei, Anhui (China); Thorne, Jonathan R.G. [Inorganic Chemistry Laboratory, Department of Chemistry, University of Oxford (United Kingdom); Tanner, Peter A. [Department of Biology and Chemistry, City University of Hong Kong, Kowloon (China)

    2002-04-15

    The direct calculation of transition line strengths and relative intensities is presented for two intraconfigurational two-photon absorption (TPA) transitions of Eu{sup 3+} in the cubic Cs{sub 2}NaYF{sub 6} host. Crystal field wavefunctions were utilized for the initial and final f{sup N}-electron states and various approaches were used in constructing all the 4f{sup N-1} 5d{sup 1} intermediate-state wavefunctions. The calculated relative intensities of the ({sup 7}F{sub 0}) {gamma}{sub 1g}{yields}({sup 5}D{sub 2}){gamma}{sub 5g}, {gamma}{sub 3g} TPA transitions are in reasonable agreement with experiment. The neglect of J-mixing in the initial state has only a small effect upon the calculation, whereas the neglect of spin-orbit couplings within the initial and terminal states drastically reduces the calculated transition linestrengths, but does not markedly change the intensity ratios. In the case of the ({sup 7}F{sub 0}){gamma}{sub 1g}{yields}({sup 5}L{sub 6}){gamma}{sub 1g}, a{gamma}{sub 5g} transitions, serious discrepancies between experiment and theory are found if the intermediate states are constructed from a 4f{sup 5} core comprising free ion states and the 5d{sup 1} crystal field states. Satisfactory agreement is, however, found when the 4f{sup 5} crystal field states are utilized in constructing the intermediate states. The contributions to the transition moment have been evaluated for various Hamiltonian terms and the results are discussed. (author)

  2. GAMSOURCE - WRS system module number 38474 for calculating gamma-ray sources produced by neutron capture

    International Nuclear Information System (INIS)

    Grimstone, M.J.

    1978-06-01

    The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the source strength of gamma-rays arising from neutron capture in a system represented in one-dimensional geometry. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)

  3. GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. N. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further

  4. Indirect probes of the trilinear Higgs coupling: $gg \\to h$ and $h \\to \\gamma \\gamma$

    CERN Document Server

    Gorbahn, Martin

    2016-10-18

    In the framework of the Standard Model effective field theory, we examine the indirect constraints on the trilinear Higgs coupling $\\lambda$ that arise from Higgs production in gluon-gluon-fusion and diphoton Higgs decays. We calculate 2-loop contributions to the $gg \\to h$ and $h \\to \\gamma \\gamma$ amplitudes that are affected by modifications of the trilinear Higgs-boson vertex. This calculation involves both the computation of anomalous dimensions and finite matching corrections. Based on our new results, we analyse the sensitivity of present and future measurements of the $hgg$ and $h \\gamma \\gamma$ couplings to shifts in $\\lambda$. Under the assumption that $O_6 = - \\lambda \\left (H^\\dagger H \\right )^3$ is the only dimension-6 operator that alters the trilinear Higgs interactions, we find that at present the considered loop-level probes provide stronger constraints than $pp \\to 2h$. At future high-energy colliders indirect ${\\cal O} (5)$ determinations of the trilinear Higgs coupling may be possible, ma...

  5. Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

    Energy Technology Data Exchange (ETDEWEB)

    Anspaugh, L.R.; Church, B.W.

    1985-12-01

    In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, Utah; Ely, Nevada; and Las Vegas, Nevada. Three events, HARRY (May 19, 1953), BEE (March 22, 1955), and SMOKY (August 31, 1957), accounted for over half of the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of ''infinite exposure,'' ''estimated exposure,'' and ''one year effective biological exposure'' are explained. 4 figs., 7 tabs.

  6. Gamma--gamma directional correlations and coincidence studies in /sup 154/Gd

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, J B; Gupta, S L; Hamilton, J H; Ramayya, A V [Vanderbilt Univ., Nashville, Tenn. (USA). Dept. of Physics; Delhi Univ. (India). Ramjas Coll.)

    1977-06-01

    The intensities, placements and E2/M1 mixing ratios of transitions in the decay of /sup 154/Eu have been carefully studied to provide accurate data for microscopic calculations. Coincidence relationships in thhe decay of /sup 154/Eu have been studied extensively with a multiparameter ..gamma..-..gamma.. coincidence system with two large volume Ge(Li) detectors. Spectra in coincidence with twenty energy gates were analyzed. Twenty-nine new coincidence relationships were established and confirmed most, but not all, of several levels previously assigned by energy fits only. From an analysis of coincidence spectra and singles spectra with a 18% efficiency Ge(Li) detector new information on the gamma-ray intensities were obtained. Precise values of the E2/M1 mixing ratios of transitions from the gamma- and beta-vibrational bands to the g.s. band have been determined from ..gamma..-..gamma.. directional correlation measurements with a NaI(Tl)-Ge(Li) detector coincidence system. Mixing ratios were obtained for a number of other transitions including those from KPI = 0/sup -/ and 2+ bands from direct and skipped cascade correlations.

  7. Gamma radiation and gamma ray protection factors of ships in various situations of radioactive fall-out

    International Nuclear Information System (INIS)

    Brehm, E.H.; Holst, T.

    1975-01-01

    In this report the development of methods of evaluating gamma ray protection factors (GSF) of ships for various situations of radioactive fall-out is described. The joining calculations of gamma ray protection factors are performed by the newly developed computer procedure GASUFA. These protection factors determine - in connection with a measured gamma radiation dose at a given detector point - the gamma radiation in different compartments of the ships. The computer program GASUFA is able to perform calculations considering the dependence of energy, place and time for the following situations: - the ship is under a radioactive cloud without fall-out; - the ship is under a radioactive cloud with fall-out; - the ship is contaminated by radioactive fall-out; - the clean or decontaminated ship is going through a zone, which is contaminated by radioactive fall-out; - the ship and the surrounding water surface are contaminated by radioactive fall-out. (orig.) [de

  8. Results of calculations of external gamma radiation exposure rates from local fallout and the related radionuclide compositions of two hypothetical 1-MT nuclear bursts. Final report

    International Nuclear Information System (INIS)

    Hicks, H.

    1984-12-01

    This report presents data on calculated gamma radiation exposure rates and local surface deposition of related radionuclides resulting from two hypothetical 1-Mt nuclear bursts. Calculations are made of the debris from two types of bombs: one containing 235 U as a fissionable material (designated oralloy), the other containing 238 U (designated tuballoy). 4 references

  9. Radio Observations of Gamma-ray Novae

    Science.gov (United States)

    Linford, Justin D.; Chomiuk, L.; Ribeiro, V.; project, E.-Nova

    2014-01-01

    Recent detection of gamma-ray emission from classical novae by the Large Area Telescope (LAT) on board the Fermi Gamma-ray Space Telescope surprised many in the astronomical community. We present results from radio observations, obtained using the Karl G. Jansky Very Large Array (VLA), of three gamma-ray novae: Mon2012, Sco2012, and Del2013. Radio observations allow for the calculation of ejecta masses, place limits on the distances, and provide information about the gamma-ray emission mechanism for these sources.

  10. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  11. The measurement of gamma ray induced heating in a mixed neutron and gamma ray environment

    International Nuclear Information System (INIS)

    Chiu, H.K.

    1991-10-01

    The problem of measuring the gamma heating in a mixed DT neutron and gamma ray environment was explored. A new detector technique was developed to make this measurement. Gamma heating measurements were made in a low-Z assembly irradiated with 14-Mev neutrons and (n, n') gammas produced by a Texas Nuclear Model 9400 neutron generator. Heating measurements were made in the mid-line of the lattice using a proportional counter operating in the Continuously-varied Bias-voltage Acquisition mode. The neutron-induced signal was separated from the gamma-induced signal by exploiting the signal rise-time differences inherent to radiations of different linear energy transfer coefficient, which are observable in a proportional counter. The operating limits of this measurement technique were explored by varying the counter position in the low-Z lattice, hence changing the irradiation spectrum observed. The experiment was modelled numerically to help interpret the measured results. The transport of neutrons and gamma rays in the assembly was modelled using the one- dimensional radiation transport code ANISN/PC. The cross-section set used for these calculations was derived from the ENDF/B-V library using the code MC 2 -2 for the case of DT neutrons slowing down in a low-Z material. The calculated neutron and gamma spectra in the slab and the relevant mass-stopping powers were used to construct weighting factors which relate the energy deposition in the counter fill-gas to that in the counter wall and in the surrounding material. The gamma energy deposition at various positions in the lattice is estimated by applying these weighting factors to the measured gamma energy deposition in the counter at those locations

  12. Neutron and gamma sensitivities of self-powered detectors: Monte Carlo modelling

    Energy Technology Data Exchange (ETDEWEB)

    Vermeeren, Ludo [SCK-CEN, Nuclear Research Centre, Boeretang 200, B-2400 Mol, (Belgium)

    2015-07-01

    This paper deals with the development of a detailed Monte Carlo approach for the calculation of the absolute neutron sensitivity of SPNDs, which makes use of the MCNP code. We will explain the calculation approach, including the activation and beta emission steps, the gamma-electron interactions, the charge deposition in the various detector parts and the effect of the space charge field in the insulator. The model can also be applied for the calculation of the gamma sensitivity of self-powered detectors and for the radiation-induced currents in signal cables. The model yields detailed information on the various contributions to the sensor currents, with distinct response times. Results for the neutron sensitivity of various types of SPNDs are in excellent agreement with experimental data obtained at the BR2 research reactor. For typical neutron to gamma flux ratios, the calculated gamma induced SPND currents are significantly lower than the neutron induced currents. The gamma sensitivity depends very strongly upon the immediate detector surroundings and on the gamma spectrum. Our calculation method opens the way to a reliable on-line determination of the absolute in-pile thermal neutron flux. (authors)

  13. Gamma ray benchmark on the spent fuel shipping cask TN 12

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Cladel, C.; Ermont, G.; Nimal, J.C.

    1983-05-01

    The purpose of this benchmark is to compare measurements and calculation of gamma-ray dose rates around a shipping cask loaded with 12 spent fuel elements of FESSENHEIM PWR type. The benchmark provides a means to verify gamma-ray sources and gamma-ray transport calculation methods in shipping cask configurations. The comparison between measurements and calculations shows a good agreement except near the fuel element top where the discrepancy reaches a factor 2

  14. NCRP Report No.151 versus Norm DIN 6847-2; NCRP Report No. 151 vs Norma DIN 6847-2

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Jimenez, J.; Rivas Ballarin, M. A.

    2008-07-01

    The National Council on Radiation protection and Measurements (NCRP) has recently published its Report No. 151, which presents recommendations and technical information on the design of structural shielding for megavoltage X and gamma-ray radiotherapy facilities. The calculations method introduced by this Report covers aspects like IMRT and other special techniques, as well as the design of structural details like doors, mazes and ducts, or the calculations of skyshine and ground shine radiation. In this work the necessary shielding for a Siemens Oncor treatment unit has been calculated, following NCRP Report No. 151 and DIN 6847-2 standard. In both cases the same isocenter workload W, use factor U, occupancy factor T and shielding design goals P, for workers and public members, are used. The results obtained with DIN 6847 are similar to the ones obtained with this Report, though there are some differences when considering the in-any-one-hour time averaged dose-equivalent rate in low occupancy factor areas, or where scattered radiation reaches the barrier under a small angle. (Author) 9 refs.

  15. Use of transmission gamma for study of calculation of incrustations thickness in oil pipelines

    International Nuclear Information System (INIS)

    Teixeira, Tâmara P.; Salgado, César M.

    2017-01-01

    Incrustations can be defined as chemical compounds organic, inorganic and mixed, initially insoluble, and which precipitate accumulating in the internal wall of pipes, surface equipment and/or parts of components involved in the production and transport of oil. These compounds, when precipitating, cause problems in the oil industry and consequently result in losses in the optimization of the extraction process. Although the importance and impact of the precipitation of these compounds in the technological and economic scope, there is still the difficulty in determining methods that enable the identification and quantification of the incrustations at an initial stage. The use of the gamma transmission technique may provide support for a better understanding of the deposition of these compounds, making it a suitable tool for the non-invasive determination of their deposition in oil transport pipelines. The geometry used for the incrustations detection include a 280 mm diameter steel pipe containing barium sulphide incrustations (BaSO 4 ) ranging from 5 to 80 mm, a gamma radiation source with divergent beam and as NaI(Tl) 2 x 2” scintillation detector. The opening size of the collimated beam was evaluated (2 to 7 mm) to also quantify the associated error in calculating the incrustations. The study was realized with computer simulation, using the MCNP-X code and validated by means of analytical equations that indicate the possibility of using this study for this purpose. (author)

  16. Use of transmission gamma for study of calculation of incrustations thickness in oil pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Tâmara P.; Salgado, César M., E-mail: tamarateixeira.eng@gmail.com, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear, (CNEN/IEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Incrustations can be defined as chemical compounds organic, inorganic and mixed, initially insoluble, and which precipitate accumulating in the internal wall of pipes, surface equipment and/or parts of components involved in the production and transport of oil. These compounds, when precipitating, cause problems in the oil industry and consequently result in losses in the optimization of the extraction process. Although the importance and impact of the precipitation of these compounds in the technological and economic scope, there is still the difficulty in determining methods that enable the identification and quantification of the incrustations at an initial stage. The use of the gamma transmission technique may provide support for a better understanding of the deposition of these compounds, making it a suitable tool for the non-invasive determination of their deposition in oil transport pipelines. The geometry used for the incrustations detection include a 280 mm diameter steel pipe containing barium sulphide incrustations (BaSO{sub 4}) ranging from 5 to 80 mm, a gamma radiation source with divergent beam and as NaI(Tl) 2 x 2” scintillation detector. The opening size of the collimated beam was evaluated (2 to 7 mm) to also quantify the associated error in calculating the incrustations. The study was realized with computer simulation, using the MCNP-X code and validated by means of analytical equations that indicate the possibility of using this study for this purpose. (author)

  17. Radiation safety aspects of the AGOR superconducting cyclotron facility

    NARCIS (Netherlands)

    Beijers, JPM; de Meijer, RJ

    1996-01-01

    This paper describes shielding calculations and skyshine estimates for the new AGOR K=600 superconducting cyclotron facility. Both simple, semi-empirical models and Monte-Carlo simulations were used. The calculations are based on a 200 MeV proton beam incident on a trick aluminum target. Also the

  18. Attenuation of the gamma rays in tissues; Atenuacion de los rayos gamma en tejidos

    Energy Technology Data Exchange (ETDEWEB)

    Arcos P, A.; Rodriguez N, S.; Pinedo S, A.; Amador V, P.; Chacon R, A.; Vega C, H.R. [Unidad Academica de Estudios Nucleares, Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2005-07-01

    The mass and lineal attenuation coefficient and of hepatic tissue, muscular, osseous and of brain before gamma rays of 10{sup -3} to 10{sup 5} MeV were calculated. For the case of the osseous tissue the calculation was made for the cartilage, the cortical tissue and the bone marrow. During the calculations the elementary composition of the tissues of human origin was used. The calculations include by separate the Photoelectric effect, the Compton scattering and the Pair production, as well as the total. For to establish a comparison with the attenuation capacities, the coefficients of the water, the aluminum and the lead also were calculated. The study was complemented measuring the attenuation coefficient of hepatic tissue of bovine before gamma rays of 0.662 MeV of a source of {sup 137} Cs. The measurement was made through of an experiment of photons transmission through samples frozen of hepatic tissue and with a Geiger-Mueller detector. (Author)

  19. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

    International Nuclear Information System (INIS)

    Lee, Yi-Kang

    2016-01-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

  20. GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M. A. [TerraPower, Bellevue, WA (United States); Lee, C. H. [TerraPower, Bellevue, WA (United States); Hill, R. N. [TerraPower, Bellevue, WA (United States)

    2017-06-28

    Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.

  1. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

  2. Survivor dosimetry. Part D. Graphical comparisons of measurements and calculations for neutrons and gamma rays

    International Nuclear Information System (INIS)

    Egbert, Stephen D.; Cullings, Harry M.

    2005-01-01

    An important part of validating the DS02 dosimetry system is the comparison of calculated initial neutron and gamma-ray radiation activation from the atomic bombs with all measurements that have been made, both before and during this current dosimetry reevaluation. All measurements that were made before the year 2002 are listed in Table 5 of Chapter 4. Many of these measurements have been compared to previous versions of the dosimetry systems for Hiroshima and Nagasaki. In this section the measurements are compared to the new dosimetry system DS02. For the purposes of showing historical context, they are also compared to the previous dosimetry system DS86. References for these measurements are found in Chapter 4. (J.P.N.)

  3. Gamma-Gompertz life expectancy at birth

    OpenAIRE

    Trifon I. Missov

    2013-01-01

    BACKGROUND The gamma-Gompertz multiplicative frailty model is the most common parametric modelapplied to human mortality data at adult and old ages. The resulting life expectancy hasbeen calculated so far only numerically. OBJECTIVE Properties of the gamma-Gompertz distribution have not been thoroughly studied. The focusof the paper is to shed light onto its first moment or, demographically speaking, characterizelife expectancy resulting from a gamma-Gompertz force of mortality. The paperprov...

  4. Energy spectrum of extragalactic gamma-ray sources

    Science.gov (United States)

    Protheroe, R. J.

    1985-01-01

    The result of Monte Carlo electron photon cascade calculations for propagation of gamma rays through regions of extragalactic space containing no magnetic field are given. These calculations then provide upper limits to the expected flux from extragalactic sources. Since gamma rays in the 10 to the 14th power eV to 10 to the 17th power eV energy range are of interest, interactions of electrons and photons with the 3 K microwave background radiation are considered. To obtain an upper limit to the expected gamma ray flux from sources, the intergalactic field is assumed to be so low that it can be ignored. Interactions with photons of the near-infrared background radiation are not considered here although these will have important implications for gamma rays below 10 to the 14th power eV if the near infrared background radiation is universal. Interaction lengths of electrons and photons in the microwave background radiation at a temperature of 2.96 K were calculated and are given.

  5. A method to describe inelastic gamma field distribution in neutron gamma density logging.

    Science.gov (United States)

    Zhang, Feng; Zhang, Quanying; Liu, Juntao; Wang, Xinguang; Wu, He; Jia, Wenbao; Ti, Yongzhou; Qiu, Fei; Zhang, Xiaoyang

    2017-11-01

    Pulsed neutron gamma density logging (NGD) is of great significance for radioprotection and density measurement in LWD, however, the current methods have difficulty in quantitative calculation and single factor analysis for the inelastic gamma field distribution. In order to clarify the NGD mechanism, a new method is developed to describe the inelastic gamma field distribution. Based on the fast-neutron scattering and gamma attenuation, the inelastic gamma field distribution is characterized by the inelastic scattering cross section, fast-neutron scattering free path, formation density and other parameters. And the contribution of formation parameters on the field distribution is quantitatively analyzed. The results shows the contribution of density attenuation is opposite to that of inelastic scattering cross section and fast-neutron scattering free path. And as the detector-spacing increases, the density attenuation gradually plays a dominant role in the gamma field distribution, which means large detector-spacing is more favorable for the density measurement. Besides, the relationship of density sensitivity and detector spacing was studied according to this gamma field distribution, therefore, the spacing of near and far gamma ray detector is determined. The research provides theoretical guidance for the tool parameter design and density determination of pulsed neutron gamma density logging technique. Copyright © 2017 Elsevier Ltd. All rights reserved.

  6. MCNP-based computational model for the Leksell gamma knife.

    Science.gov (United States)

    Trnka, Jiri; Novotny, Josef; Kluson, Jaroslav

    2007-01-01

    We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large

  7. Dose analysis on high performance vault storage system of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Kei-ichiro; Maki, Koichi; Shimizu, Masashi; Oda, Masashi; Kumagai, Naoki [Hitachi, Ltd., Power and Industrial Systems R and D Laboratory, Hitachi, Ibaraki (Japan); Hoshikawa, Tadahiro; Oyama, Kenichi; Kanai, Hidetoshi [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    2000-03-01

    The radiation shielding design for the high performance vault storage system is studied in order to satisfy the design targets at the controlled area and the site boundaries. The additional gamma-ray shields in front of the storage tubes and the shielding structures at the entrance of the ducts are installed. The dose rates at the inlet and the outlet of the ducts are estimated by 3D calculation. And the dose rates at the controlled area and the site boundaries are evaluated taking the effect of the direct radiation and the indirect one (skyshine) into consideration. The dose rates at the controlled area and the site boundaries are about 7x10{sup -7} Sv/h and 3x10{sup -10} Sv/h, respectively. Thus, we have the prospect to satisfy the design targets. (author)

  8. In-situ gamma spectrometry method for determination of environmental gamma dose; Metodo de espectrometria gamma in situ para determinacao de dose gama ambiental

    Energy Technology Data Exchange (ETDEWEB)

    Conti, Claudio de Carvalho

    1995-07-15

    This work tries to establish a methodology for germanium detectors calibration, normally used for in situ gamma ray spectrometry, for determining the environmental exposure rate in function of the energy of the incident photons. For this purpose a computer code has been developed, based on the stripping method, for the computational spectra analysis to calculate the contribution of the partial absorption of the gamma rays (Compton effect) in the active and nonactive parts of the detector. The resulting total absorption spectrum is then converted to fluence distribution in function of the energy for the photons reaching the detector, which is then used to calculate the exposure rate or kerma in air. The unfolding and fluency convention parameters are determined by detector calibration using point gamma sources. The method is validated by comparison of the results against the calculated exposure rate at a point of interest for the standards. This method is used for the direct measurement of the exposure rate distribution in function of the energy at the site, in situ measurement technic, leading to rapid results during an emergency situation and also used for indoor measurements. (author)

  9. Gamma response study of radiation sensitive MOSFETs for their use as gamma radiation sensor

    Energy Technology Data Exchange (ETDEWEB)

    Srivastava, Saurabh; Kumar, A. Vinod [Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai (India); Aggarwal, Bharti; Singh, Arvind; Topkar, Anita, E-mail: anita@barc.gov.in [Electronics Division, Bhabha Atomic Research Centre, Mumbai (India)

    2016-05-23

    Continuous monitoring of gamma dose is important in various fields like radiation therapy, space-related research, nuclear energy programs and high energy physics experiment facilities. The present work is focused on utilization of radiation-sensitive Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs) to monitor gamma radiation doses. Static characterization of these detectors was performed to check their expected current-voltage relationship. Threshold voltage and transconductance per unit gate to source voltage (K factor) were calculated from the experimental data. The detector was exposed to gamma radiation in both, with and without gate bias voltage conditions, and change in threshold voltage was monitored at different gamma doses. The experimental data was fitted to obtain equation for dependence of threshold voltage on gamma dose. More than ten times increase in sensitivity was observed in biased condition (+3 V) compared to the unbiased case.

  10. GammaWorkshops Proceedings

    DEFF Research Database (Denmark)

    Strålberg, Elisabeth; Klemola, Seppo; Nielsen, Sven Poul

    to the GammaWorkshops. The topics included efficiency transfer, true coincidence summing corrections, self-attenuation corrections, measurement of natural radionuclides (natural decay series), combined measurement uncertainty calculations, and detection limits. These topics covered both lectures and practical...

  11. Skyshine - a paper tiger

    International Nuclear Information System (INIS)

    Rindi, A.; Thomas, R.H.

    1975-01-01

    The study of the transport to large distances of radiation produced by high-energy accelerators is of fundamental interest and important in the calculation of the exposure of the population living in their environment. The most significant of the experimental data accumulated since the construction of the early particle accelerators and their theoretical interpretation are reviewed. (author)

  12. A new measurement of the rare decay eta -> pi^0 gamma gamma with the Crystal Ball/TAPS detectors at the Mainz Microtron

    Energy Technology Data Exchange (ETDEWEB)

    Nefkens, B M; Prakhov, S; Aguar-Bartolom��, P; Annand, J R; Arends, H J; Bantawa, K; Beck, R; Bekrenev, V; Bergh��user, H; Braghieri, A; Briscoe, W J; Brudvik, J; Cherepnya, S; Codling, R F; Collicott, C; Costanza, S; Danilkin, I V; Denig, A; Demissie, B; Dieterle, M; Downie, E J; Drexler, P; Fil' kov, L V; Fix, A; Garni, S; Glazier, D I; Gregor, R; Hamilton, D; Heid, E; Hornidge, D; Howdle, D; Jahn, O; Jude, T C; Kashevarov, V L; K��ser, A; Keshelashvili, I; Kondratiev, R; Korolija, M; Kotulla, M; Koulbardis, A; Kruglov, S; Krusche, B; Lisin, V; Livingston, K; MacGregor, I J; Maghrbi, Y; Mancel, J; Manley, D M; McNicoll, E F; Mekterovic, D; Metag, V; Mushkarenkov, A; Nikolaev, A; Novotny, R; Oberle, M; Ortega, H; Ostrick, M; Ott, P; Otte, P B; Oussena, B; Pedroni, P; Polonski, A; Robinson, J; Rosner, G; Rostomyan, T; Schumann, S; Sikora, M H; Starostin, A; Strakovsky, I I; Strub, T; Suarez, I M; Supek, I; Tarbert, C M; Thiel, M; Thomas, A; Unverzagt, M; Watts, D P; Werthmueller, D; Witthauer, L

    2014-08-01

    A new measurement of the rare, doubly radiative decay eta->pi^0 gamma gamma was conducted with the Crystal Ball and TAPS multiphoton spectrometers together with the photon tagging facility at the Mainz Microtron MAMI. New data on the dependence of the partial decay width, Gamma(eta->pi^0 gamma gamma), on the two-photon invariant mass squared, m^2(gamma gamma), as well as a new, more precise value for the decay width, Gamma(eta->pi^0 gamma gamma) = (0.33+/-0.03_tot) eV, are based on analysis of 1.2 x 10^3 eta->pi^0 gamma gamma decays from a total of 6 x 10^7 eta mesons produced in the gamma p -> eta p reaction. The present results for dGamma(eta->pi^0 gamma gamma)/dm^2(gamma gamma) are in good agreement with previous measurements and recent theoretical calculations for this dependence.

  13. High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

    International Nuclear Information System (INIS)

    Yasin, Zafar; Matei, Catalin; Ur, Calin A.; Mitu, Iani-Octavian; Udup, Emil; Petcu, Cristian

    2016-01-01

    The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKA and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.

  14. High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

    Energy Technology Data Exchange (ETDEWEB)

    Yasin, Zafar, E-mail: zafar.yasin@eli-np.ro; Matei, Catalin; Ur, Calin A.; Mitu, Iani-Octavian; Udup, Emil; Petcu, Cristian [Extreme Light Infrastructure - Nuclear Physics / Horia Hulubei National Institute for R& D in Physics and Nuclear Engineering, Bucharest-Magurele (Romania)

    2016-03-25

    The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKA and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.

  15. Gamma factors of an ambulatory source; Factores gamma de una fuente ambulatoria

    Energy Technology Data Exchange (ETDEWEB)

    Arcos P, A; Vega C, H R; Manzanares A, E; Salas L, M A; Hernandez D, V M [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Barquero, R [Hospital Universitario del Rio Hortega, E-47010 Valladolid (Spain)

    2007-07-01

    Some of the procedures for diagnostic or treatment used in the medicine use radioactive materials as the I{sup 131}. By means of Monte Carlo methods were calculated the doses in the internal organs of a woman, with three months of pregnancy, due to the radioiodine captured by her thyroid, as well as to 1 meter of the gland. A three-dimensional mathematical model of the body of a woman was used and by means of Monte Carlo, the radioiodine photons were transported isotropically from the thyroid toward the whole body and was calculated the absorbed dose by their internal organs, also the Kerma in air (K) was determined and the environmental equivalent dose (H{sup *}(10)) at 1 m of the gland. Two activity factors at dose were determined, Gamma Factors that it allows to estimate the dose that the patient produces to people to its around. Of the gamma radiation that emits the I{sup 131} in the thyroid was found that the thymus receives the biggest dose while the uterus is the organ that smaller dose receives. The determined gamma factors were: {gamma}{sub KAire} = 56 {mu}Gy-m{sup 2}-h{sup -1}-GBq{sup -1}, and {gamma}{sub H}{sup *}{sub (10)} = 73 {mu}Sv-m{sup 2}-h{sup -1}-GBq{sup -1}. The distribution of the absorbed dose by the internal organs is attributed to the relative distance among the thyroid and the other organs, to the inter-organs shielding, its size and to its elementary composition. The {gamma}{sub KAire} and {gamma}{sub H}{sup *}{sub (10)} factors allow to estimate the exposure that the patient produces on the personnel to its around. With this, the nuclear medicus, the medical physicist or the one responsible of the radiological safety in the hospital can give more precise indications on the behavior of people around the patient. (Author)

  16. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

    2016-11-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

  17. Comparison of penumbra regions produced by ancient Gamma knife model C and Gamma ART 6000 using Monte Carlo MCNP6 simulation.

    Science.gov (United States)

    Banaee, Nooshin; Asgari, Sepideh; Nedaie, Hassan Ali

    2018-07-01

    The accuracy of penumbral measurements in radiotherapy is pivotal because dose planning computers require accurate data to adequately modeling the beams, which in turn are used to calculate patient dose distributions. Gamma knife is a non-invasive intracranial technique based on principles of the Leksell stereotactic system for open deep brain surgeries, invented and developed by Professor Lars Leksell. The aim of this study is to compare the penumbra widths of Leksell Gamma Knife model C and Gamma ART 6000. Initially, the structure of both systems were simulated by using Monte Carlo MCNP6 code and after validating the accuracy of simulation, beam profiles of different collimators were plotted. MCNP6 beam profile calculations showed that the penumbra values of Leksell Gamma knife model C and Gamma ART 6000 for 18, 14, 8 and 4 mm collimators are 9.7, 7.9, 4.3, 2.6 and 8.2, 6.9, 3.6, 2.4, respectively. The results of this study showed that since Gamma ART 6000 has larger solid angle in comparison with Gamma Knife model C, it produces better beam profile penumbras than Gamma Knife model C in the direct plane. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Monte Carlo simulations of plutonium gamma-ray spectra

    International Nuclear Information System (INIS)

    Koenig, Z.M.; Carlson, J.B.; Wang, Tzu-Fang; Ruhter, W.D.

    1993-01-01

    Monte Carlo calculations were investigated as a means of simulating the gamma-ray spectra of Pu. These simulated spectra will be used to develop and evaluate gamma-ray analysis techniques for various nondestructive measurements. Simulated spectra of calculational standards can be used for code intercomparisons, to understand systematic biases and to estimate minimum detection levels of existing and proposed nondestructive analysis instruments. The capability to simulate gamma-ray spectra from HPGe detectors could significantly reduce the costs of preparing large numbers of real reference materials. MCNP was used for the Monte Carlo transport of the photons. Results from the MCNP calculations were folded in with a detector response function for a realistic spectrum. Plutonium spectrum peaks were produced with Lorentzian shapes, for the x-rays, and Gaussian distributions. The MGA code determined the Pu isotopes and specific power of this calculated spectrum and compared it to a similar analysis on a measured spectrum

  19. Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-07-01

    This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

  20. Study of the performance of a 4 {pi} {gamma} ionisation chamber; Etude des performances d'une chambre d'ionisation 4 {pi} {gamma}

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    Description of a well-type {gamma} chamber for routine measurements of sources between 1 {mu}curie and 1 curie. Approximate calculation of some geometrical parameters by which the zone of constant sensitivity may be increased. Calculation of the sensitivity of such a chamber and comparison with the experimental results. (author) [French] Description d'une chambre a puits {gamma}, destinee aux mesures de routine de sources entre 1 {mu}curie et 1 curie. Calcul approche de quelques parametres geometriques permettant d'accroitre la zone d'egale sensibilite. Calcul de la sensibilite d'une telle chambre et confrontation avec les resultats experimentaux. (auteur)

  1. Gamma spectrometry and plastic-scintillator inherent background

    International Nuclear Information System (INIS)

    Pomerantsev, V.V.; Gagauz, I.B.; Mitsai, L.I.; Pilipenko, V.S.; Solomonov, V.M.; Chernikov, V.V.; Tsirlin, Y.A.

    1988-01-01

    The authors measured the energy resolution for a linear dependence of light yield on gamma radiation energy of gamma spectrometers based on plastic scintillation detectors for several plastic scintillators. If there were several gamma lines from the source the line with the highest energy was used to eliminate distortion due to overlap from the Compton background from gamma radiation of higher energy. Attenuation lengths were calculated. The tests were based on three modes of interaction between the gamma radiation and the scintillator: Compton scattering, the photoelectric effect, and pair formation. The contribution from light collection was also considered. The scintillators tested included polystyrene, polymethyl methacrylate, cesium iodide, and sodium iodide. Gamma sources included cesium 137, sodium 22, potassium 40, yttrium 88, thorium 232, and plutonium-beryllium

  2. Effect of intermediate zone during gamma-gamma and X-ray spectrum logging of quarry wells with inversive probes

    International Nuclear Information System (INIS)

    Artsybashev, V.A.; Volkov, A.A.

    1981-01-01

    Experimental and calculated data on the effect of intermediate zone on results of gamma-gamma and X-ray spectrum logging of quarry wells at chalcopyrite deposits are presented. The measurements have been carried out using 57 Co and 109 Cd sources for intermediate zone represented by water, air and mud. Requirements for technical conditions of wells have been formulated. According to the requirements the application of gamma- gamma method is possible when the thickness of the intermediate zone does not exceed several millimetres and that of X-ray spectrum method - when the thickness does not exceed several hundredths of a millimetre [ru

  3. The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.

    2004-01-01

    An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)

  4. Fluorescence of the gamma, epsilon, and delta systems of nitric oxide - Polarization and use of calculated intensities for spectrometer calibration.

    Science.gov (United States)

    Poland, H. M.; Broida, H. P.

    1971-01-01

    Results of a study in which fluorescence of the gamma system of nitric oxide was obtained by excitation from both the 2144 A line of ionized cadmium and a continuum source. Individual rotational lines of the 2144 A excited fluorescence spectrum were found to be partially polarized and to have polarizations of differ ing sign. Measured relative vibrational band intensities from line and continuum excitation were compared to calculated Franck-Condon factors. Those Franck-Condon factors based on a single potential for the two spin states of the X super pi state agreed better with measured values than those based on separate potentials for the two spin states. Calculated intensities of the v prime = 3 progression were used to calibrate the instrument response in the wavelength region from 2000 to 2500 A and were checked with measured intensities of the v prime = 0.1, and 2 progressions. Fluorescence of the epsilon and delta bands obtained with continuum lamp excitation also were compared to calculated intensities.

  5. Ford motor company NDE facility shielding design

    International Nuclear Information System (INIS)

    Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

  6. Ford Motor Company NDE facility shielding design.

    Science.gov (United States)

    Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

  7. Gamma-ray Output Spectra from 239 Pu Fission

    International Nuclear Information System (INIS)

    Ullmann, John

    2015-01-01

    Gamma-ray multiplicities, individual gamma-ray energy spectra, and total gamma energy spectra following neutron-induced fission of 239 Pu were measured using the DANCE detector at Los Alamos. Corrections for detector response were made using a forward-modeling technique based on propagating sets of gamma rays generated from a paramaterized model through a GEANT model of the DANCE array and adjusting the parameters for best fit to the measured spectra. The results for the gamma-ray spectrum and multiplicity are in general agreement with previous results, but the measured total gamma-ray energy is about 10% higher. A dependence of the gamma-ray spectrum on the gamma-ray multplicity was also observed. Global model calculations of the multiplicity and gamma energy distributions are in good agreement with the data, but predict a slightly softer total-energy distribution

  8. Monte Carlo neutron and gamma-ray calculations

    International Nuclear Information System (INIS)

    Mendelsohn, Edgar

    1987-01-01

    Kerma in tissue and the activation produced in sulfur and cobalt due to prompt neutrons from the Hiroshima and Nagasaki bombs were calculated out to 2000 m from the hypocenter in 100 m increments. As neutron sources weapon output spectra calculated by investigators from the Los Alamos National Laboratory (LANL) were used. Other parameters, such as burst height and air and ground densities and compositions, were obtained from recent sources. The LLNL Monte Carlo transport code TART was used for these calculations. TART accesses the well-established 1985 ENDL cross-section library, which has built-in reaction cross sections. The zoning for this problem was a full two-dimensional geometry with a ceiling height of 1100 m and a ground thickness of 30 cm. For the Hiroshima calculations (including sulfur activation) and untilted source was used. However, a special sulfur activation problem using a source tilted 15 deg was run for which the ratios to the untilted case are reported. The TART code uses a technique for solving the transport equation that is different from that of the ORNL DOT code; it also draws on a specially evaluated cross-section library (ENDL) and uses a larger group structure than DOT. One of the purposes of this work was to instill confidence in the DOT calculations that will be used directly in the dose reassessment of A-bomb survivors. The TART results were compared with values calculated with the DOT code by investigators from ORNL and found to be in good agreement for the most part. However, the sulfur activation comparison is disappointing. Because the sulfur activation is caused by higher energy neutrons (which should have experienced fewer collisions than those causing cobalt activation, for example), better agreement than what is reported here would be expected

  9. DIGA/NSL new calculational model in slab geometry

    International Nuclear Information System (INIS)

    Makai, M.; Gado, J.; Kereszturi, A.

    1987-04-01

    A new calculational model is presented based on a modified finite-difference algorithm, in which the coefficients are determined by means of the so-called gamma matrices. The DIGA program determines the gamma matrices and the NSL program realizes the modified finite difference model. Both programs assume slab cell geometry, DIGA assumes 2 energy groups and 3 diffusive regions. The DIGA/NSL programs serve to study the new calculational model. (author)

  10. Text book of dose calculation for operators

    International Nuclear Information System (INIS)

    Aoyagi, Haruki; Gonda, Kozo

    1979-07-01

    This is a text book of dose calculation for the operators of the reprocessing factory of Power Reactor and Nuclear Fuel Development Corporation. The radiations considered are beta-ray and gamma-ray. The method used is a point attenuation nuclear integral method. Radiation sources are considered as the assemblies of point sources. Dose from each point source is calculated, then, total dose is obtained by the integration for all sources. Attenuation is calculated by considering the attenuation owing to distance and the absorption by absorbers. The build-up factor is introduced for the correction for scattered gamma-ray. The build-up factor is given in a table for various scatterers. The operators are able to calculate dose by themselves. The results of integral calculation expressed with formulas are given in graphs. (Kato, T.)

  11. Motor power calculation for driving conveyor chain in gamma irradiator BATAN 2x250 k curie

    International Nuclear Information System (INIS)

    Ari Satmoko; Syamsurrijal Ramdja; Sutomo Budihardjo

    2010-01-01

    Recently, an Irradiator BATAN 2X250 k Curie for agricultural product is under design. The installation is provided by the gamma source about 2x250 k Curie. Agricultural products are carried into carriers and these carriers are hanged on the conveyor chain. The chain moves into a radiation chamber following the trajectoire. The chain is drived by motor. For this reason, the calculation is performed to determine the motor power. After resolving the force equilibrium equation, the force and power of the motor needed to drive the chain are obtained. Numerical method by using V Basic language is used to resolve the equation. The calculation result shows the correlation between friction coefficient and motor power. From the evaluation, it is decided that the friction coefficient should be less than 0,015. By this friction, the motor power is about 3. 13 k Watt. From the evaluation, it is also obtained that the radius of the curve trajectory shall not be too small. Combination between high friction and small curve radius could lead to the locked condition in which high power motor are not be able to move the conveyor chain). (author)

  12. Validity test of design calculations of a PGNAA setup

    International Nuclear Information System (INIS)

    Naqvi, A.A.; Garwan, M.A.

    2004-01-01

    A rectangular moderator has been designed for the prompt gamma ray neutron activation analysis (PGNAA) setup at King Fahd University of Petroleum and Minerals (KFUPM) to analyze Portland cement samples. The design of the moderator assembly was obtained using Monte Carlo calculations. The design calculations of the new rectangular moderator of the KFUPM PGNAA setup have been verified experimentally through prompt gamma ray yield measurement as a function of the front moderator thickness. In this study the yield of the 3.54 and 4.94 MeV prompt gamma rays from silicon in a soil sample was measured as a function of thickness of the front moderator of the rectangular moderator. The experimental results were compared with the results of the Monte Carlo simulations. A good agreement has been achieved between the experimental results and the results of the calculations. The experimental results have provided useful information about the PGNAA setup performance, neutron moderation, and gamma ray attenuation in the PGNAA sample

  13. Relative branching ratio of the {eta}{yields}{pi}{sup 0}{gamma}{gamma} decay channel

    Energy Technology Data Exchange (ETDEWEB)

    Knecht, N.; Papandreou, Z.; Lolos, G.J.; Benslama, K.; Huber, G.M.; Li, S.; Bekrenev, V.; Briscoe, W.J.; Grosnick, D.; Isenhower, D.; Koetke, D.D.; Kozlenko, N.G.; Kruglov, S.; Manley, D.M.; Manweiler, R.; McDonald, S.; Olmsted, J.; Shafi, A.; Stanislaus, T.D.S

    2004-06-03

    The {eta}{yields}{pi}{sup 0}{gamma}{gamma} rare decay was measured at the AGS with the Crystal Ball photon spectrometer and its relative branching ratio was extracted to be B{sub 1}=(8.3{+-}2.8{+-}1.4)x10{sup -4}, based on the analysis of 3x10{sup 7} detected {eta} mesons. This leads to a lower partial width for this eta channel than past measurements and is in line with chiral perturbation theory calculations.

  14. Novel Radiobiological Gamma Index for Evaluation of 3-Dimensional Predicted Dose Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Sumida, Iori, E-mail: sumida@radonc.med.osaka-u.ac.jp [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Yamaguchi, Hajime; Kizaki, Hisao; Aboshi, Keiko; Tsujii, Mari; Yoshikawa, Nobuhiko; Yamada, Yuji [Department of Radiation Oncology, NTT West Osaka Hospital, Osaka (Japan); Suzuki, Osamu; Seo, Yuji [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Isohashi, Fumiaki [Department of Radiation Oncology, NTT West Osaka Hospital, Osaka (Japan); Yoshioka, Yasuo [Department of Radiation Oncology, Osaka University Graduate School of Medicine, Osaka (Japan); Ogawa, Kazuhiko [Department of Radiation Oncology, NTT West Osaka Hospital, Osaka (Japan)

    2015-07-15

    Purpose: To propose a gamma index-based dose evaluation index that integrates the radiobiological parameters of tumor control (TCP) and normal tissue complication probabilities (NTCP). Methods and Materials: Fifteen prostate and head and neck (H&N) cancer patients received intensity modulated radiation therapy. Before treatment, patient-specific quality assurance was conducted via beam-by-beam analysis, and beam-specific dose error distributions were generated. The predicted 3-dimensional (3D) dose distribution was calculated by back-projection of relative dose error distribution per beam. A 3D gamma analysis of different organs (prostate: clinical [CTV] and planned target volumes [PTV], rectum, bladder, femoral heads; H&N: gross tumor volume [GTV], CTV, spinal cord, brain stem, both parotids) was performed using predicted and planned dose distributions under 2%/2 mm tolerance and physical gamma passing rate was calculated. TCP and NTCP values were calculated for voxels with physical gamma indices (PGI) >1. We propose a new radiobiological gamma index (RGI) to quantify the radiobiological effects of TCP and NTCP and calculate radiobiological gamma passing rates. Results: The mean RGI gamma passing rates for prostate cases were significantly different compared with those of PGI (P<.03–.001). The mean RGI gamma passing rates for H&N cases (except for GTV) were significantly different compared with those of PGI (P<.001). Differences in gamma passing rates between PGI and RGI were due to dose differences between the planned and predicted dose distributions. Radiobiological gamma distribution was visualized to identify areas where the dose was radiobiologically important. Conclusions: RGI was proposed to integrate radiobiological effects into PGI. This index would assist physicians and medical physicists not only in physical evaluations of treatment delivery accuracy, but also in clinical evaluations of predicted dose distribution.

  15. Lambda-guided calculation method (LGC method) for xenon/CT CBF

    Energy Technology Data Exchange (ETDEWEB)

    Sase, Shigeru [Anzai Medical Co., Ltd., Tokyo (Japan); Honda, Mitsuru; Kushida, Tsuyoshi; Seiki, Yoshikatsu; Machida, Keiichi; Shibata, Iekado [Toho Univ., Tokyo (Japan). School of Medicine

    2001-12-01

    A quantitative CBF calculation method for xenon/CT was developed by logically estimating time-course change rate (rate constant) of arterial xenon concentration from that of end-tidal xenon concentration. A single factor ({gamma}) was introduced to correlate the end-tidal rate constant (Ke) with the arterial rate constant (Ka) in a simplified equation. This factor ({gamma}) is thought to reflect the diffusing capacity of the lung for xenon. When an appropriate value is given to {gamma}, it is possible to calculate the arterial rate constant (Calculated Ka) from Ke. To determine {gamma} for each xenon/CT CBF examination, a procedure was established which utilizes the characteristics of white matter lambda; lambda refers to xenon brain-blood partition coefficient. Xenon/CT studies were performed on four healthy volunteers. Hemispheric CBF values (47.0{+-}9.0 ml/100 g/min) with use of Calculated Ka were close to the reported normative values. For a 27-year-old healthy man, the rate constant for the common carotid artery was successfully measured and nearly equal to Calculated Ka. The authors conclude the method proposed in this work, lambda-guided calculation method, could make xenon/CT CBF substantially reliable and quantitative by effective use of end-tidal xenon. (author)

  16. Steady-State Core Temperature Prediction Based on GAMMA+/CAPP Coupling

    International Nuclear Information System (INIS)

    Tak, Nam-il; Lee, Hyun-Chul; Lim, Hong-Sik

    2015-01-01

    In spite of sizable applications of the GAMMA+ code for the thermo-fluid analysis and design of a prismatic VHTR, the existing works are limited to stand-alone calculations. In the stand-alone calculations, information from the neutronic analysis (e.g., reactor power density profile) was considered only once i.e., when the calculations get started. For the neutronic analysis and design of a VHTR, the CAPP code, which is also under development at KAERI, is used. The main objective of this paper is to investigate the capability of GAMMA+ and CAPP coupling and to examine the results of the coupled analysis. Based on the coupling of GAMMA+ and CAPP, the steady-state core temperature was investigated in this work. It is found that the communication of data was successful. And the results of the GAMMA+ and CAPP coupling are found to be reasonable. The design modification of PMR200 is required to satisfy the design limit for the hot spot fuel temperature

  17. Thermoluminescence of nanocrystalline CaSO{sub 4}: Dy for gamma dosimetry and calculation of trapping parameters using deconvolution method

    Energy Technology Data Exchange (ETDEWEB)

    Mandlik, Nandkumar, E-mail: ntmandlik@gmail.com [Department of Physics, University of Pune, Ganeshkhind, Pune -411007, India and Department of Physics, Fergusson College, Pune- 411004 (India); Patil, B. J.; Bhoraskar, V. N.; Dhole, S. D. [Department of Physics, University of Pune, Ganeshkhind, Pune -411007 (India); Sahare, P. D. [Department of Physics and Astrophysics, University of Delhi, Delhi- 110007 (India)

    2014-04-24

    Nanorods of CaSO{sub 4}: Dy having diameter 20 nm and length 200 nm have been synthesized by the chemical coprecipitation method. These samples were irradiated with gamma radiation for the dose varying from 0.1 Gy to 50 kGy and their TL characteristics have been studied. TL dose response shows a linear behavior up to 5 kGy and further saturates with increase in the dose. A Computerized Glow Curve Deconvolution (CGCD) program was used for the analysis of TL glow curves. Trapping parameters for various peaks have been calculated by using CGCD program.

  18. A gamma beam profile imager for ELI-NP Gamma Beam System

    Science.gov (United States)

    Cardarelli, P.; Paternò, G.; Di Domenico, G.; Consoli, E.; Marziani, M.; Andreotti, M.; Evangelisti, F.; Squerzanti, S.; Gambaccini, M.; Albergo, S.; Cappello, G.; Tricomi, A.; Veltri, M.; Adriani, O.; Borgheresi, R.; Graziani, G.; Passaleva, G.; Serban, A.; Starodubtsev, O.; Variola, A.; Palumbo, L.

    2018-06-01

    The Gamma Beam System of ELI-Nuclear Physics is a high brilliance monochromatic gamma source based on the inverse Compton interaction between an intense high power laser and a bright electron beam with tunable energy. The source, currently being assembled in Magurele (Romania), is designed to provide a beam with tunable average energy ranging from 0.2 to 19.5 MeV, rms energy bandwidth down to 0.5% and flux of about 108 photons/s. The system includes a set of detectors for the diagnostic and complete characterization of the gamma beam. To evaluate the spatial distribution of the beam a gamma beam profile imager is required. For this purpose, a detector based on a scintillator target coupled to a CCD camera was designed and a prototype was tested at INFN-Ferrara laboratories. A set of analytical calculations and Monte Carlo simulations were carried out to optimize the imager design and evaluate the performance expected with ELI-NP gamma beam. In this work the design of the imager is described in detail, as well as the simulation tools used and the results obtained. The simulation parameters were tuned and cross-checked with the experimental measurements carried out on the assembled prototype using the beam from an x-ray tube.

  19. Multiple Gamma-Ray Detection Capability of a CeBr3 Detector for Gamma Spectroscopy

    Directory of Open Access Journals (Sweden)

    A. A. Naqvi

    2017-01-01

    Full Text Available The newly developed cerium tribromide (CeBr3 detector has reduced intrinsic gamma-ray activity with gamma energy restricted to 1400–2200 keV energy range. This narrower region of background gamma rays allows the CeBr3 detector to detect more than one gamma ray to analyze the gamma-ray spectrum. Use of multiple gamma-ray intensities in elemental analysis instead of a single one improves the accuracy of the estimated results. Multigamma-ray detection capability of a cylindrical 75 mm × 75 mm (diameter × height CeBr3 detector has been tested by analyzing the chlorine concentration in water samples using eight chlorine prompt gamma rays over 517 to 8578 keV energies utilizing a D-D portable neutron generator-based PGNAA setup and measuring the corresponding minimum detection limit (MDC of chlorine. The measured MDC of chlorine for gamma rays with 517–8578 keV energies varies from 0.07 ± 0.02 wt% to 0.80 ± 0.24. The best value of MDC was measured to be 0.07 ± 0.02 wt% for 788 keV gamma rays. The experimental results are in good agreement with Monte Carlo calculations. The study has shown excellent detection capabilities of the CeBr3 detector for eight prompt gamma rays over 517–8578 keV energy range without significant background interference.

  20. Tank Z-361 dose rate calculations

    International Nuclear Information System (INIS)

    Richard, R.F.

    1998-01-01

    Neutron and gamma ray dose rates were calculated above and around the 6-inch riser of tank Z-361 located at the Plutonium Finishing Plant. Dose rates were also determined off of one side of the tank. The largest dose rate 0.029 mrem/h was a gamma ray dose and occurred 76.2 cm (30 in.) directly above the open riser. All other dose rates were negligible. The ANSI/ANS 1991 flux to dose conversion factor for neutrons and photons were used in this analysis. Dose rates are reported in units of mrem/h with the calculated uncertainty shown within the parentheses

  1. Determination of gamma ray shielding parameters of rocks and concrete

    Science.gov (United States)

    Obaid, Shamsan S.; Gaikwad, Dhammajyot K.; Pawar, Pravina P.

    2018-03-01

    Gamma shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff) and electron density (Neff) have been measured and calculated for rocks and concrete in the energy range 122-1330 keV. The measurements have been carried out at 122, 356, 511, 662, 1170, 1275, 1330 keV gamma ray energies using a gamma spectrometer includes a NaI(Tl) scintillation detector and MCA card. The atomic and electronic cross sections have also been investigated. Experimental and calculated (WinXCom) values were compared, and good agreement has been observed within the experimental error. The obtained results showed that feldspathic basalt, compact basalt, volcanic rock, dolerite and pink granite are more efficient than the sandstone and concrete for gamma ray shielding applications.

  2. Response functions of NaI(Tl) detectors to terrestrial gamma radiation

    International Nuclear Information System (INIS)

    Gyurcsak, J.; Lenda, A.

    1978-01-01

    Computer programs, serving for calculation of detector efficiency and energy deposition spectrum for scintillation crystals irradiated by isotropic or half-isotropic gamma-ray fields were elaborated. The Monte-Carlo models used in calculations are valid for gamma-ray energies 2 π geometry by the 1.5'' x 2'' probe with experimental results is given. (author)

  3. Born order study of {gamma}{sup *}{gamma}{sup *} {yields} {rho}{rho} at very high energy

    Energy Technology Data Exchange (ETDEWEB)

    Pire, B. [Ecole Polytechnique, 91 - Palaiseau (France). Centre de Physique Theorique; Szymanowski, L. [Soltan Institute for Nuclear Studies, Warsaw (Poland); Liege Univ. (Belgium); Wallon, S. [Paris-11 Univ., Lab. de Physique Theorique, 91 - Orsay (France)

    2005-07-01

    We calculate the cross-section for the diffractive exclusive process {gamma}{sub L}{sup *}(Q{sub 1}{sup 2}){gamma}{sub L}{sup *}(Q{sub 2}{sup 2}) {yields} {rho}{sub L}{sup 0}{rho}{sub L}{sup 0}, in view of its study in the future high energy e{sup +}e{sup -} linear collider. The Born order approximation of the amplitude is completely calculable in the hard region Q{sub 1}{sup 2},Q{sub 2}{sup 2} >> {lambda}{sup 2}(QCD). The resulting cross-section is large enough for this process to be measurable with foreseen luminosity and energy, for Q{sub 1}{sup 2} and Q{sub 2}{sup 2} in the range of a few GeV{sup 2}. (authors)

  4. Gamma multi-detectors and nuclear structure studies: search for superdeformed structures in {sup 147}Gd and {sup 144}Gd isotopes using Crystal Castle; simulation calculations for EUROGAM multi-detector definition; Multidetecteurs gamma et etudes de structure nucleaire: recherche avec le Chateau de Cristal de structures superdeformees dans les isotopes {sup 147}Gd et {sup 144}Gd; calculs de simulation pour la definition du multidetecteur EUROGAM

    Energy Technology Data Exchange (ETDEWEB)

    France, G de

    1992-12-31

    Computer simulations have been used for the calculation of the new generation of 4 {pi}{gamma} multi-detectors (Castle Crystal) of EUROGAM system (phase I and II). Two superdeformed bands (I and II), comprising 16 and 13 transitions respectively, have been described for {sup 147}Gd nucleus during the {sup 122}Sn({sup 30}Si,5n) fusion-evaporation reaction in a 155 MeV bombardment energy. Dynamic inertia momentum similarities and gamma transition energy similarities have been observed between band I and {sup 148}Gd nucleus and between band II and {sup 146}Gd nucleus, respectively. These similarities can be related to a pseudo-spin symmetry. Calculations suggest the existence of an octupolar susceptibility in this mass region. {sup 144}Gd nucleus has been studied using {sup 120}Sn({sup 29}Si,5n) fusion-evaporation reaction in a 155 MeV bombardment energy and using {sup 100}Mo({sup 48}Ti,4n) reactions in a 200 MeV bombardment energy. {gamma}-{gamma} coincidences have revealed the existence of a 58 keV width valley in the matrix representation compatible with theoretical predictions. In spite of the evidence for about ten transitions during these experiments, no superdeformed structure has been demonstrated for {sup 144}Gd nucleus. (J.S.). 87 refs., 57 figs., 41 tabs.

  5. The transport of neutrons and gamma-rays in the air

    International Nuclear Information System (INIS)

    Adamski, J.

    1980-01-01

    The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)

  6. Gamma-ray emission cross section from proton-incident spallation reaction

    International Nuclear Information System (INIS)

    Iga, Kiminori; Ishibashi, Kenji; Shigyo, Nobuhiro

    1996-01-01

    Gamma-ray emission double differential cross sections from proton-incident spallation reaction have been measured at incident energies of 0.8, 1.5 and 3.0 GeV with Al, Fe, In and Pb targets. The experimental results have been compared with calculate values of HETC-KFA2. The measured cross sections disagree with the calculated results in the gamma ray energies above 10 MeV. (author)

  7. Gamma-ray energy buildup factor calculations and shielding effects of some Jordanian building structures

    Science.gov (United States)

    Sharaf, J. M.; Saleh, H.

    2015-05-01

    The shielding properties of three different construction styles, and building materials, commonly used in Jordan, were evaluated using parameters such as attenuation coefficients, equivalent atomic number, penetration depth and energy buildup factor. Geometric progression (GP) method was used to calculate gamma-ray energy buildup factors of limestone, concrete, bricks, cement plaster and air for the energy range 0.05-3 MeV, and penetration depths up to 40 mfp. It has been observed that among the examined building materials, limestone offers highest value for equivalent atomic number and linear attenuation coefficient and the lowest values for penetration depth and energy buildup factor. The obtained buildup factors were used as basic data to establish the total equivalent energy buildup factors for three different multilayer construction styles using an iterative method. The three styles were then compared in terms of fractional transmission of photons at different incident photon energies. It is concluded that, in case of any nuclear accident, large multistory buildings with five layers exterior walls, style A, could effectively attenuate radiation more than small dwellings of any construction style.

  8. Attenuation of the gamma rays in tissues

    International Nuclear Information System (INIS)

    Arcos P, A.; Rodriguez N, S.; Pinedo S, A.; Amador V, P.; Chacon R, A.; Vega C, H.R.

    2005-01-01

    The mass and lineal attenuation coefficient and of hepatic tissue, muscular, osseous and of brain before gamma rays of 10 -3 to 10 5 MeV were calculated. For the case of the osseous tissue the calculation was made for the cartilage, the cortical tissue and the bone marrow. During the calculations the elementary composition of the tissues of human origin was used. The calculations include by separate the Photoelectric effect, the Compton scattering and the Pair production, as well as the total. For to establish a comparison with the attenuation capacities, the coefficients of the water, the aluminum and the lead also were calculated. The study was complemented measuring the attenuation coefficient of hepatic tissue of bovine before gamma rays of 0.662 MeV of a source of 137 Cs. The measurement was made through of an experiment of photons transmission through samples frozen of hepatic tissue and with a Geiger-Mueller detector. (Author)

  9. $\\gamma^{*}\\gamma^{*}$ total cross-section in the dipole picture of BFKL dynamics

    CERN Document Server

    Boonekamp, M; Royon, C; Wallon, S

    1999-01-01

    The total $\\gamma^*\\gamma^*$ cross-section is derived in the Leading Order QCD dipole picture of BFKL dynamics, and compared with the one from 2-gluon exchange. The Double Leading Logarithm approximation of the DGLAP cross-section is found to be small in the phase space studied. Cross sections are calculated for realistic data samples at the $e^+e^-$ collider LEP and a future high energy linear collider. Next to Leading order corrections to the BFKL evolution have been determined phenomenologically, and are found to give very large corrections to the BFKL cross-section, leading to a reduced sensitivity for observing BFKL.

  10. Influence of the rock heterogeneity on the results of gamma-gamma logging

    International Nuclear Information System (INIS)

    Umiastowski, K.; Buniak, M.

    1977-01-01

    The influence of the grain size of granular rock and that of the empty holes size of porous rock on the results of gamma-gamma measurements was investigated. Monte-Carlo calculations and experiments were performed to establish this influence. For the grain (or empty holes) size greater than about 10mm the significant influence on the results of density measurements was found if a 137 Cs source was used. This effect is greater for the dry rock than for the water-saturated one. Formulae enabling the correct density value to be found, if the grain size is known, were proposed. (author)

  11. Double diffractive {rho} -production in {gamma}{sup *}{gamma}{sup *} collisions

    Energy Technology Data Exchange (ETDEWEB)

    Pire, B. [Ecole Polytechnique, CPHT, Palaiseau (France); Szymanowski, L. [Soltan Institute for Nuclear Studies, Warsaw (Poland); Universite de Liege, Liege (Belgium); Wallon, S. [Universite Paris-Sud, LPT, Orsay (France)

    2005-12-01

    We present a first estimate of the cross-section for the exclusive process {gamma}{sup *}{sub L}(Q{sub 1}{sup 2}){gamma}{sup *}{sub L}(Q{sub 2}{sup 2}){yields}{rho}{sub L}{sup 0}{rho}{sub L}{sup 0}, which will be studied in the future high energy e{sup +} e{sup -}-linear collider. As a first step, we calculate the Born order approximation of the amplitude for longitudinally polarized virtual photons and mesons, in the kinematical region s >>-t, Q{sub 1}{sup 2}, Q{sub 2}{sup 2}. This process is completely calculable in the hard region Q{sub 1}{sup 2}, Q{sub 2}{sup 2}>>{lambda}{sup 2}{sub QCD}. We perform most of the steps in an analytical way. The resulting cross-section turns out to be large enough for this process to be measurable with foreseen luminosity and energy, for Q{sub 1}{sup 2} and Q{sub 2}{sup 2} in the range of a few GeV{sup 2}. (orig.)

  12. Characteristics of a gamma telescope on the ''Kosmos-561'' satellite

    International Nuclear Information System (INIS)

    Bokov, V.L.; Kruglov, E.M.

    1981-01-01

    The results of calculations of gamma telescope characteristics intended for investigating cosmic γ radiation at E>=100 MeV in the ''Cosmos 561'' artificial Earth satellite, using the Monte Carlo method, are presented. The gamma spectrometer contains a lead converter, scintillation deteectors of polysterene, a unit of spark chambers and a Cherenkov detector of lead glass. The dependence of the device effective area and angular resolution on γ quanta energy is calculated. The relative radiation pattern of the device is given. The given integral characteristics of the gamma telescope for a γ quanta flux with an exponential spectrum are the following: the effective geometrical factor and effective device area depending on the spectrum index. The calibration gamma telescope curve is plotted according to the electron mean free path distribution [ru

  13. Dose mapping simulation using the MCNP code for the Syrian gamma irradiation facility and benchmarking

    International Nuclear Information System (INIS)

    Khattab, K.; Boush, M.; Alkassiri, H.

    2013-01-01

    Highlights: • The MCNP4C was used to calculate the gamma ray dose rate spatial distribution in for the SGIF. • Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter was conducted as well. • Good agreements were noticed between the calculated and measured results. • The maximum relative differences were less than 7%, 4% and 4% in the x, y and z directions respectively. - Abstract: A three dimensional model for the Syrian gamma irradiation facility (SGIF) is developed in this paper to calculate the gamma ray dose rate spatial distribution in the irradiation room at the 60 Co source board using the MCNP-4C code. Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter is conducted as well to compare the calculated and measured results. Good agreements are noticed between the calculated and measured results with maximum relative differences less than 7%, 4% and 4% in the x, y and z directions respectively. This agreement indicates that the established model is an accurate representation of the SGIF and can be used in the future to make the calculation design for a new irradiation facility

  14. A method for determination mass absorption coefficient of gamma rays by Compton scattering

    International Nuclear Information System (INIS)

    El Abd, A.

    2014-01-01

    A method was proposed for determination mass absorption coefficient of gamma rays for compounds, alloys and mixtures. It is based on simulating interaction processes of gamma rays with target elements having atomic numbers from Z=1 to Z=92 using the MCSHAPE software. Intensities of Compton scattered gamma rays at saturation thicknesses and at a scattering angle of 90° were calculated for incident gamma rays of different energies. The obtained results showed that the intensity of Compton scattered gamma rays at saturations and mass absorption coefficients can be described by mathematical formulas. These were used to determine mass absorption coefficients for compound, alloys and mixtures with the knowledge of their Compton scattered intensities. The method was tested by calculating mass absorption coefficients for some compounds, alloys and mixtures. There is a good agreement between obtained results and calculated ones using WinXom software. The advantages and limitations of the method were discussed. - Highlights: • Compton scattering of γ−rays was used for determining mass absorption coefficient. • Scattered intensities were determined by the MCSHAPE software. • Mass absorption coefficients were determined for some compounds, mixtures and alloys. • Mass absorption coefficients were calculated by Winxcom software. • Good agreements were found between determined and calculated results

  15. SOILD: A computer model for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil

    International Nuclear Information System (INIS)

    Chen, S.Y.; LePoire, D.; Yu, C.; Schafetz, S.; Mehta, P.

    1991-01-01

    The SOLID computer model was developed for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil. It is designed to assess external doses under various exposure scenarios that may be encountered in environmental restoration programs. The models four major functional features address (1) dose versus source depth in soil, (2) shielding of clean cover soil, (3) area of contamination, and (4) nonuniform distribution of sources. The model is also capable of adjusting doses when there are variations in soil densities for both source and cover soils. The model is supported by a data base of approximately 500 radionuclides. 4 refs

  16. A method for determination mass absorption coefficient of gamma rays by Compton scattering.

    Science.gov (United States)

    El Abd, A

    2014-12-01

    A method was proposed for determination mass absorption coefficient of gamma rays for compounds, alloys and mixtures. It is based on simulating interaction processes of gamma rays with target elements having atomic numbers from Z=1 to Z=92 using the MCSHAPE software. Intensities of Compton scattered gamma rays at saturation thicknesses and at a scattering angle of 90° were calculated for incident gamma rays of different energies. The obtained results showed that the intensity of Compton scattered gamma rays at saturations and mass absorption coefficients can be described by mathematical formulas. These were used to determine mass absorption coefficients for compound, alloys and mixtures with the knowledge of their Compton scattered intensities. The method was tested by calculating mass absorption coefficients for some compounds, alloys and mixtures. There is a good agreement between obtained results and calculated ones using WinXom software. The advantages and limitations of the method were discussed. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. Electronic, elastic, thermodynamic properties and structure disorder of {gamma}-AlON solid solution from ab initio calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yuezhong, E-mail: wyzphysics@163.com [Department of Physics and Key Laboratory for Radiation Physics and Technology of Ministry of Education, Sichuan University, Chengdu 610064 (China); Tianjin Jinhang Institute of Technical Physics, Tianjin 300192 (China); Lu, Tiecheng, E-mail: lutiecheng@scu.edu.cn [Department of Physics and Key Laboratory for Radiation Physics and Technology of Ministry of Education, Sichuan University, Chengdu 610064 (China); International Center for Material Physics, Chinese Academy of Sciences, Shenyang 110015 (China); Zhang, Rongshi [Tianjin Jinhang Institute of Technical Physics, Tianjin 300192 (China); Jiang, Shengli; Qi, Jianqi; Wang, Ying [Department of Physics and Key Laboratory for Radiation Physics and Technology of Ministry of Education, Sichuan University, Chengdu 610064 (China); Chen, Qingyun [Department of Physics and Key Laboratory for Radiation Physics and Technology of Ministry of Education, Sichuan University, Chengdu 610064 (China); National Defense Key Discipline Laboratory of Nuclear Waste and Environmental Safety, Southwest University of Science and Technology, Mianyang 621010 (China); Miao, Naihua [Physique Theorique des Materiaux, Universite de Liege, Sart Tilman B-4000 (Belgium); He, Duanwei [Institute of Atomic and Molecular Physics, Sichuan University, Chengdu 610064 (China)

    2013-01-25

    Highlights: Black-Right-Pointing-Pointer We reassess the chemical bonding character of {gamma}-AlON which shows strong ionicity. Black-Right-Pointing-Pointer {gamma}-AlON single-crystals exhibit highly elastic anisotropy. Black-Right-Pointing-Pointer The thermodynamic properties are investigated in a wider temperature/pressure range. Black-Right-Pointing-Pointer {gamma}-AlON is an O/N partially disordered structure. - Abstract: Spinel aluminium oxynitride ({gamma}-AlON), as a kind of transparent ceramic material expectable, is studied using the ab initio density functional method, in terms of electronic, elastic, thermodynamic properties and structure disorder. The results show that {gamma}-AlON exhibits strong ionicity, as quantitatively expressed by (Al{sub O}{sup 2.43+}){sub 15}(Al{sub T}{sup 2.41+}){sub 8}(O{sup 1.64-}){sub 27}(N{sup 2.27-}){sub 5} from our reassessment of the ionic character. We summarize and speculate that the considered oxynitride single-crystals exhibit highly elastic anisotropy. The interpretation of the thermodynamic properties of {gamma}-AlON according to quasi-harmonic Debye model confirm the available experiments and are extended to a wider temperature/pressure range. This material holds high elastic strength under extreme environments, where dB/dT absolute value is less than 0.03 GPa/K, independent of the pressure. Finally, we study the O/N structure disorder character of {gamma}-AlON solid solution by investigating nine possible crystal structures. It is found that {gamma}-AlON should be partially disordered, and in fact, the O/N ordering has a significant effect on the properties.

  18. On the omnipresent background gamma radiation of the continuous spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Banjanac, R.; Maletić, D.; Joković, D., E-mail: yokovic@ipb.ac.rs; Veselinović, N.; Dragić, A.; Udovičić, V.; Aničin, I.

    2014-05-01

    The background spectrum of a germanium detector, shielded from the radiations arriving from the lower and open for the radiations arriving from the upper hemisphere, is studied by means of absorption measurements, both in a ground level and in an underground laboratory. The low-energy continuous portion of this background spectrum that peaks at around 100 keV, which is its most intense component, is found to be of very similar shape at the two locations. It is established that it is mostly due to the radiations of the real continuous spectrum, which is quite similar to the instrumental one. The intensity of this radiation is in our cases estimated to about 8000 photons/(m{sup 2}s·2π·srad) in the ground level laboratory, and to about 5000 photons/(m{sup 2}s·2π·srad) in the underground laboratory, at the depth of 25 m.w.e. Simulations by GEANT4 and CORSIKA demonstrate that this radiation is predominantly of terrestrial origin, due to environmental gamma radiations scattered off the materials that surround the detector (the “skyshine radiation”), and to a far less extent to cosmic rays of degraded energy. - Highlights: • We studied the low-energy part of continuous background spectra of germanium detectors. • The study was performed at the ground level and at the shallow underground sites. • The instrumental spectrum is due to radiations of the similar continuous spectrum. • The low-energy radiation is of both terrestrial and cosmic-ray origin. • In our study, we find that this radiation is of predominantly terrestrial origin.

  19. THIDA: code system for calculation of the exposure dose rate around a fusion device

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Igarashi, Masahito.

    1978-12-01

    A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

  20. Gamma ray lines from a universal extra dimension

    Energy Technology Data Exchange (ETDEWEB)

    Bertone, Gianfranco; Jackson, C. B.; Shaughnessy, Gabe; Tait, Tim M.P.; Vallinotto, Alberto

    2012-03-01

    Indirect Dark Matter searches are based on the observation of secondary particles produced by the annihilation or decay of Dark Matter. Among them, gamma-rays are perhaps the most promising messengers, as they do not suffer deflection or absorption on Galactic scales, so their observation would directly reveal the position and the energy spectrum of the emitting source. Here, we study the detailed gamma-ray energy spectrum of Kaluza--Klein Dark Matter in a theory with 5 Universal Extra Dimensions. We focus in particular on the two body annihilation of Dark Matter particles into a photon and another particle, which produces monochromatic photons, resulting in a line in the energy spectrum of gamma rays. Previous calculations in the context of the five dimensional UED model have computed the line signal from annihilations into \\gamma \\gamma, but we extend these results to include \\gamma Z and \\gamma H final states. We find that these spectral lines are subdominant compared to the predicted \\gamma \\gamma signal, but they would be important as follow-up signals in the event of the observation of the \\gamma \\gamma line, in order to distinguish the 5d UED model from other theoretical scenarios.

  1. Gamma-Gompertz life expectancy at birth

    Directory of Open Access Journals (Sweden)

    Trifon I. Missov

    2013-02-01

    Full Text Available BACKGROUND The gamma-Gompertz multiplicative frailty model is the most common parametric modelapplied to human mortality data at adult and old ages. The resulting life expectancy hasbeen calculated so far only numerically. OBJECTIVE Properties of the gamma-Gompertz distribution have not been thoroughly studied. The focusof the paper is to shed light onto its first moment or, demographically speaking, characterizelife expectancy resulting from a gamma-Gompertz force of mortality. The paperprovides an exact formula for gamma-Gompertz life expectancy at birth and a simplerhigh-accuracy approximation that can be used in practice for computational convenience.In addition, the article compares actual (life-table to model-based (gamma-Gompertzlife expectancy to assess on aggregate how many years of life expectancy are not captured(or overestimated by the gamma-Gompertz mortality mechanism. COMMENTS A closed-form expression for gamma-Gomeprtz life expectancy at birth contains a special(the hypergeometric function. It aids assessing the impact of gamma-Gompertz parameterson life expectancy values. The paper shows that a high-accuracy approximation canbe constructed by assuming an integer value for the shape parameter of the gamma distribution.A historical comparison between model-based and actual life expectancy forSwedish females reveals a gap that is decreasing to around 2 years from 1950 onwards.Looking at remaining life expectancies at ages 30 and 50, we see this gap almost disappearing.

  2. Spectra of {gamma} rays feeding superdeformed bands

    Energy Technology Data Exchange (ETDEWEB)

    Lauritsen, T.; Khoo, T.L.; Henry, R.G. [and others

    1995-08-01

    The spectrum of {gamma}rays coincident with SD transitions contains the transitions which populate the SD band. This spectrum can provide information on the feeding mechanism and on the properties (moment of inertia, collectivity) of excited SD states. We used a model we developed to explain the feeding of SD bands, to calculate the spectrum of feeding {gamma}rays. The Monte Carlo simulations take into account the trigger conditions present in our Eurogam experiment. Both experimental and theoretical spectra contain a statistical component and a broad E2 peak (from transitions occurring between excited states in the SD well). There is good resemblance between the measured and calculated spectra although the calculated multiplicity of an E2 bump is low by {approximately}30%. Work is continuing to improve the quality of the fits, which will result in a better understanding of excited SD states. In addition, a model for the last steps, which cool the {gamma} cascade into the SD yrast line, needs to be developed. A strong M1/E2 low-energy component, which we believe is responsible for this cooling, was observed.

  3. Alternative methodology for irradiation reactor experimental shielding calculation

    International Nuclear Information System (INIS)

    Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

    1996-01-01

    Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

  4. Source and replica calculations

    International Nuclear Information System (INIS)

    Whalen, P.P.

    1994-01-01

    The starting point of the Hiroshima-Nagasaki Dose Reevaluation Program is the energy and directional distributions of the prompt neutron and gamma-ray radiation emitted from the exploding bombs. A brief introduction to the neutron source calculations is presented. The development of our current understanding of the source problem is outlined. It is recommended that adjoint calculations be used to modify source spectra to resolve the neutron discrepancy problem

  5. The determination and use of radionuclide background in gamma spectrometry

    International Nuclear Information System (INIS)

    Zimmer, W.H.

    1986-01-01

    Background is the major component of gross photon peak area. Therefore, net area, nuclide activity, counting uncertainty, and limits of detection calculations are no better than the calculation of background. In this study, background in gamma spectrometry is explored in several of its aspects. Means are presented to reduce background. Standard practices are presented to be used in the acquisition of valid, relevant background data. Unified standard calculations with examples are presented in the use of background data to determine net count and counting uncertainty. L. A. Currie's latest calculations of Lower Limits of Detection (1) (LLD) as they apply to gamma spectrometry are reviewed. Finally, Maximum Undetected Activity (MUA), LLD, and Critical Level (CL) concepts and calculations are compared in sample spectra

  6. Presentation of a semiempirical method for the calculation of doses due to neutrons and capture gamma rays inside high energy accelerators rooms

    International Nuclear Information System (INIS)

    Larcher, A.M.; Bonet Duran, S.M.

    1998-01-01

    Full text: Medical electron accelerators operating above 10 MeV produce radiation beams that are contaminated with neutrons. Therefore, shielding design for high energy accelerator rooms must consider the neutron component of the radiation field. In this paper a semiempirical method is presented to calculate doses due to neutrons and capture gamma rays inside the room and the maze. The calculation method is based on the knowledge of the neutron yield Q (neutrons/Gy of photons at isocenter) and the average energy of the primary beam of neutrons Eo (MeV). The method constitutes an appropriate tool for shielding facilities evaluation. The accuracy of the method has been contrasted with data obtained from the literature and an excellent correlation among the calculations and the measured values was achieved. In addition, the method has been used in the verification of experimental data corresponding to a 15 MeV linear accelerator installed in the country with similar results. (author) [es

  7. Mathematical simulation of gamma-radiation angle distribution measurements

    International Nuclear Information System (INIS)

    Batij, V.G.; Batij, E.V.; Egorov, V.V.; Fedorchenko, D.V.; Kochnev, N.A.

    2008-01-01

    We developed mathematical model of the facility for gamma-radiation angle distribution measurement and calculated response functions for gamma-radiation intensities. We developed special software for experimental data processing, the 'Shelter' object radiation spectra unfolding and Sphere detector (ShD) angle resolution estimation. Neuronet method using for detection of the radiation directions is given. We developed software based on the neuronet algorithm, that allows obtaining reliable distribution of gamma-sources that make impact on the facility detectors at the measurement point. 10 refs.; 15 figs.; 4 tab

  8. Polycrystalline Materials as a Cold Neutron and Gamma Radiation Filter

    International Nuclear Information System (INIS)

    Habib, N.

    2009-01-01

    The total neutron cross-section of polycrystalline beryllium, graphite and iron has been calculated beyond their cut-off wavelength using a general formula. The computer Cold Filter code was developed in order to provide the required calculations. The code also permits the calculation of attenuation of reactor gamma radiation, The calculated neutron transmissions through polycrystalline Be graphite and iron at different temperatures were compared with the experimental data measured at the ETRR-1 reactor using two TOF spectrometers. An overall agreement is obtained between the formula fits and experimental data at different temperatures. A feasibility study is carried on using polycrystalline Be, graphite and iron an efficient filter for cold neutrons and gamma radiation.

  9. Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

    International Nuclear Information System (INIS)

    Pudjijanto, M.S.; Akhmad, Y.R.

    1998-01-01

    A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

  10. Positron imaging with multiwire proportional chamber-gamma converter hybrid detectors

    International Nuclear Information System (INIS)

    Chu, D.Y.H.

    1976-09-01

    A large area positron camera was developed using multiwire proportional chambers as detectors and electromagnetic delay lines for coordinate readout. Honeycomb structured gamma converters made of lead are coupled to the chambers for efficient gamma detection and good spatial resolution. Two opposing detectors, each having a sensitive area of 48 cm x 48 cm, are operated in coincidence for the detection of annihilation gammas (511 keV) from positron emitters. Detection efficiency of 4.2 percent per detector and spatial resolution of 6 to 7 mm FWHM at the mid-plane were achieved. The present camera operates at a maximum count rate of 24 K counts/min, limited by accidental coincidence. The theory for the gamma converter is presented along with a review of the operation of the multiwire proportional chamber and delay line readout. Calculated gamma converter efficiencies are compared with the measured results using a prototype test chamber. The characteristics of the positron camera system is evaluated, and the performance is shown to be consistent with calculation

  11. Calculation of the BREN house shielding experiments

    International Nuclear Information System (INIS)

    Woolson, William A.; Gritzner, Michael L.

    1987-01-01

    The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

  12. An automated Monte-Carlo based method for the calculation of cascade summing factors

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, M.J., E-mail: mark.j.jackson@awe.co.uk; Britton, R.; Davies, A.V.; McLarty, J.L.; Goodwin, M.

    2016-10-21

    A versatile method has been developed to calculate cascade summing factors for use in quantitative gamma-spectrometry analysis procedures. The proposed method is based solely on Evaluated Nuclear Structure Data File (ENSDF) nuclear data, an X-ray energy library, and accurate efficiency characterisations for single detector counting geometries. The algorithm, which accounts for γ–γ, γ–X, γ–511 and γ–e{sup −} coincidences, can be applied to any design of gamma spectrometer and can be expanded to incorporate any number of nuclides. Efficiency characterisations can be derived from measured or mathematically modelled functions, and can accommodate both point and volumetric source types. The calculated results are shown to be consistent with an industry standard gamma-spectrometry software package. Additional benefits including calculation of cascade summing factors for all gamma and X-ray emissions, not just the major emission lines, are also highlighted. - Highlights: • Versatile method to calculate coincidence summing factors for gamma-spectrometry analysis. • Based solely on ENSDF format nuclear data and detector efficiency characterisations. • Enables generation of a CSF library for any detector, geometry and radionuclide. • Improves measurement accuracy and reduces acquisition times required to meet MDA.

  13. Calculation of dose conversion factors for doses in the fingernails to organ doses at external gamma irradiation in air

    International Nuclear Information System (INIS)

    Khailov, A.M.; Ivannikov, A.I.; Skvortsov, V.G.; Stepanenko, V.F.; Orlenko, S.P.; Flood, A.B.; Williams, B.B.; Swartz, H.M.

    2015-01-01

    Absorbed doses to fingernails and organs were calculated for a set of homogenous external gamma-ray irradiation geometries in air. The doses were obtained by stochastic modeling of the ionizing particle transport (Monte Carlo method) for a mathematical human phantom with arms and hands placed loosely along the sides of the body. The resulting dose conversion factors for absorbed doses in fingernails can be used to assess the dose distribution and magnitude in practical dose reconstruction problems. For purposes of estimating dose in a large population exposed to radiation in order to triage people for treatment of acute radiation syndrome, the calculated data for a range of energies having a width of from 0.05 to 3.5 MeV were used to convert absorbed doses in fingernails to corresponding doses in organs and the whole body as well as the effective dose. Doses were assessed based on assumed rates of radioactive fallout at different time periods following a nuclear explosion. - Highlights: • Elemental composition and density of nails were determined. • MIRD-type mathematical human phantom with arms and hands was created. • Organ doses and doses to nails were calculated for external photon exposure in air. • Effective dose and nail doses values are close for rotational and soil surface exposures.

  14. Absorbed dose calculation from beta and gamma rays of 131I in ellipsoidal thyroid and other organs of neck with MCNPX code

    Directory of Open Access Journals (Sweden)

    Mohammad Mirzaie

    2012-09-01

    Full Text Available Background: The 131I radioisotope is used for diagnosis and treatment of hyperthyroidism and thyroid cancer. In optimized Iodine therapy, a specific dose must be reached to the thyroid gland with minimum radiation to the cervical spine, cervical vertebrae, neck tissue, subcutaneous fat and skin. Dose measurement inside the alive organ is difficult therefore the aim of this research was dose calculation in the organs by MCNPX code. Materials and Methods: First of all, the input file for MCNPX code has been prepared to calculate F6 and F8 tallies for ellipsoidal thyroid lobes with long axes is tow times of short axes which the 131I is distributed uniformly inside the lobes. Then the code has been run for F6 and F8 tallies for variation of lobe volume from 1 to 25 milliliters. From the output file of tally F6, the gamma absorbed dose in ellipsoidal thyroid, spinal neck, neck bone, neck tissue, subcutaneous fat layer and skin for the volume lobe variation from 1 ml to 25 ml have been derived and the graphs are drew. As well as, form the output of F8 tally the absorbed energy of beta in thyroid and soft tissue of neck is obtained and listed in the table and then absorbed dose of bate has been calculated. Results: The results of this research show that for constant activity in thyroid, the absorbed dose of gamma decreases about 88.3% in thyroid, 6.9% at soft tissue, 19.3% in adipose layer and 17.4% in skin, but it increases 32.1% in spinal of neck and 32.3% in neck bone when the lobe volume varied from 1 to 25 milliliters. For the same situation, the beta absorbed dose decreases 95.9% in thyroid and 64.2% in soft tissue. Conclusion: For the constant activity in thyroid by increasing the thyroid volume, absorbed dose of gamma in thyroid and soft tissue of neck, adipose layer under the skin and skin of neck decreased, but it increased at spinal of neck and neck bone. Also, by increasing of the lobe volume in constant activity, the beta absorbed dose

  15. Evaluation of the computer code system RADHEAT-V4 by analysing benchmark problems on radiation shielding

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Naito, Yoshitaka

    1990-11-01

    A computer code system RADHEAT-V4 has been developed for safety evaluation on radiation shielding of nuclear fuel facilities. To evaluate the performance of the code system, 18 benchmark problem were selected and analysed. Evaluated radiations are neutron and gamma-ray. Benchmark problems consist of penetration, streaming and skyshine. The computed results show more accurate than those by the Sn codes ANISN and DOT3.5 or the Monte Carlo code MORSE. Big core memory and many times I/O are, however, required for RADHEAT-V4. (author)

  16. Spectrometric gamma radiation of shale cores applied to sweet spot discrimination in Eastern Pomerania, Poland

    Science.gov (United States)

    Skupio, Rafal; de Alemar Barberes, Gabriel

    2017-12-01

    This paper describes the application and calculation of hydrocarbon anomalies in two different boreholes located in Eastern Pomerania (northern Poland). Spectrometric data from borehole geophysical probe (borehole 1) and portable gamma logger (borehole 2) were used to analyze shale formations. The results from borehole 1 presented a statistically significant, moderate correlation between calculated hydrocarbon anomalies and hydrocarbon saturation data obtained from well log interpretation. Borehole 2 has been analyzed focusing on the gamma radiation of the core samples, and the positive results of borehole 1. Hydrocarbon anomalies calculated from spectral gamma radiation are reliable indicators of sweet spots, based solely on a cursory evaluation of core measurements. These preliminary information acquired from gamma-ray measurements could help increase sampling precision of further geochemical analysis.

  17. Planetary gamma-ray spectroscopy: the effects of hydrogen absorption cross-section of the gamma-ray spectrum

    International Nuclear Information System (INIS)

    Lapides, J.R.

    1981-01-01

    The gamma-ray spectroscopy of planet surfaces is one of several possible methods that are useful in determining the elemental composition of planet surfaces from orbiting spacecraft. This has been demonstrated on the Apollos 15 and 16 missions as well as the Soviet Mars-5 mission. Planetary gamma-ray emission is primarily the result of natural radioactive decay and cosmic-ray and solar-flare-induced nuclear reactions. Secondary neutron reactions play a large role in the more intense gamma-ray emission. The technique provides information on the elemental composition of the top few tens of centimeters of the planet surface. Varying concentrations of hydrogen and compositional variations that alter the macroscopic thermal-neutron absorption cross section have a significant effect on the neutron flux in the planet surface and therefore also on the gamma-ray emission from the surface. These effects have been systematically studied for a wide range of possible planetary compositions that include Mercury, the moon, Mars, the comets, and the asteroids. The problem of the Martian atmosphere was also investigated. The results of these calculations, in which both surface neutron fluxes and gamma-ray emission fluxes were determined, were used to develop general procedures for obtaining planet compositions from the gamma-ray spectrum. Several changes have been suggested for reanalyzing the Apollos 15 and 16 gamma-ray results. In addition, procedures have been suggested that can be applied to neutron-gamma techniques in mineral and oil exploration

  18. QCD factorizations in {gamma}*{gamma}*->{rho}{sub L}{sup 0}{rho}{sub L}{sup 0}

    Energy Technology Data Exchange (ETDEWEB)

    Pire, B. [CPHT, Unite mixte 7644 du CNRS, Ecole Polytechnique, 91128 Palaiseau (France)]. E-mail: pire@cpht.polytechnique.fr; Segond, M. [LPT, Unite mixte 8627 du CNRS, Universite Paris-Sud, 91405 Orsay (France); Szymanowski, L. [LPT, Unite mixte 8627 du CNRS, Universite Paris-Sud, 91405 Orsay (France); Universite de Liege, B-4000 Liege (Belgium); Soltan Institute for Nuclear Studies, Hoza 69, 00-681 Warsaw (Poland); Wallon, S. [LPT, Unite mixte 8627 du CNRS. , Universite Paris-Sud, 91405 Orsay (France)

    2006-08-24

    We calculate the lowest order QCD amplitude, i.e. the quark exchange contribution, to the forward production amplitude of a pair of longitudinally polarized {rho} mesons in the scattering of two virtual photons {gamma}*(Q{sub 1}){gamma}*(Q{sub 2})->{rho}{sub L}{sup 0}{rho}{sub L}{sup 0}. We show that the scattering amplitude simultaneously factorizes in two quite different ways: the part with transverse photons is described by the QCD factorization formula involving the generalized distribution amplitude of two final {rho} mesons, whereas the part with longitudinally polarized photons takes the QCD factorized form with the {gamma}{sub L}*->{rho}{sub L}{sup 0} transition distribution amplitude. Perturbative expressions for these, in general, non-perturbative functions are obtained in terms of the {rho}-meson distribution amplitude.

  19. Selection of skin dose calculation methodologies

    International Nuclear Information System (INIS)

    Farrell, W.E.

    1987-01-01

    This paper reports that good health physics practice dictates that a dose assessment be performed for any significant skin contamination incident. There are, however, several methodologies that could be used, and while there is probably o single methodology that is proper for all cases of skin contamination, some are clearly more appropriate than others. This can be demonstrated by examining two of the more distinctly different options available for estimating skin dose the calculational methods. The methods compiled by Healy require separate beta and gamma calculations. The beta calculational method is the derived by Loevinger, while the gamma dose is calculated from the equation for dose rate from an infinite plane source with an absorber between the source and the detector. Healy has provided these formulas in graphical form to facilitate rapid dose rate determinations at density thicknesses of 7 and 20 mg/cm 2 . These density thicknesses equate to the regulatory definition of the sensitive layer of the skin and a more arbitrary value to account of beta absorption in contaminated clothing

  20. An automated Monte-Carlo based method for the calculation of cascade summing factors

    Science.gov (United States)

    Jackson, M. J.; Britton, R.; Davies, A. V.; McLarty, J. L.; Goodwin, M.

    2016-10-01

    A versatile method has been developed to calculate cascade summing factors for use in quantitative gamma-spectrometry analysis procedures. The proposed method is based solely on Evaluated Nuclear Structure Data File (ENSDF) nuclear data, an X-ray energy library, and accurate efficiency characterisations for single detector counting geometries. The algorithm, which accounts for γ-γ, γ-X, γ-511 and γ-e- coincidences, can be applied to any design of gamma spectrometer and can be expanded to incorporate any number of nuclides. Efficiency characterisations can be derived from measured or mathematically modelled functions, and can accommodate both point and volumetric source types. The calculated results are shown to be consistent with an industry standard gamma-spectrometry software package. Additional benefits including calculation of cascade summing factors for all gamma and X-ray emissions, not just the major emission lines, are also highlighted.

  1. Characteristics of liver tissue for attenuate the gamma radiation; Caracteristicas del tejido hepatico para atenuar la radiacion gamma

    Energy Technology Data Exchange (ETDEWEB)

    Arcos P, A.; Rodriguez N, S.; Pinedo S, A.; Amador V, P.; Chacon R, A.; Vega C, H.R. [Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2005-07-01

    It was determined the lineal attenuation coefficient of hepatic tissue before gamma radiation of a source of {sup 137} Cs. When exposing organic material before X or gamma radiation fields, part of the energy of the photons is absorbed by the material, while another part crosses it without producing any effect. The quantity of energy that is absorbed is a measure of the dose that receives the material. The three main mechanisms by means of which the gamma rays interacting with the matter are: The Photoelectric Effect, the Compton dispersion and the Even production; the sum of these three processes is translated in the attenuation coefficient of the radiation. In this work we have used hepatic tissue of bovine, as substitute of the human hepatic tissue, and we have measured the lineal attenuation coefficient for photons of 0.662 MeV. Through a series of calculations we have determined the lineal attenuation coefficient for photons from 10{sup -3} to 10{sup -5} MeV and the measured coefficient was compared with the one calculated. (Author)

  2. Environmental Gamma Radiation Measurements in Baskil District

    International Nuclear Information System (INIS)

    Canbazoglu, C.

    2008-01-01

    In this study, we have determined environmental gamma radiation dose rate in Baskil district which has very high granite content in its geographical structure. Gamma radiation dose rate measurements were achieved by portable radiation monitoring equipment based on the energy range between 40 keV and 1.3 MeV. The measurements were performed on asphalt and soil surface level and also one meter above the ground surface. The gamma dose rate was also performed inside and outside of buildings over the district. The dose rates were found to be between 8.46μR/h and 34.66 μR/h. Indoor and outdoor effective dose rate of the gamma radiation exposure has been calculated to be 523μSv/y and 196μSv/y, respectively

  3. Apparatus for gamma radiography

    International Nuclear Information System (INIS)

    1983-06-01

    The aim of the present standard is to fix the rules for the construction of gamma radiography instrumentation without prejudice to the present regulations. These apparatus have to be fitted with only sealed sources conformable to the experimental standard M 61-002. The present standard agrees with the international standard ISO 3999 of 1977 dealing with the same subject. Nevertheless, it is different on the three main following points: it does not accept the same limits of absorbed dose rates in the air calculated on the external surface of projectors; it precribes tightness, bending, crushing and tensile tests for some components of the gamma radiography it prescribes tests of endurance and resistance to breaking for the locking systems of the gamma radiography apparatus. The present standard also specifies the following points: symbols and indications to put on projectors and on the source-holder; identification of the source contained in the projector; and, accompanying documents. The regulation references are given in annexe [fr

  4. The self-absorption effect of gamma rays in 239Pu

    International Nuclear Information System (INIS)

    Hsiaohua Hsu

    1989-01-01

    Nuclear materials assay with gamma-ray spectrum measurement is a well-established method for safeguards. However, for a thick source, the self-absorption of characteristic low-energy gamma rays has been a handicap to accurate assay. The author has carried out Monte Carlo simulations to study this effect using the 239 Pu α-decay gamma-ray spectrum as an example. The thickness of a plutonium metal source can be considered a function of gamma-ray intensity ratios. In a practical application, gamma-ray intensity ratios can be obtained from a measured spectrum. With the help of calculated curves, scientists can find the source thickness and make corrections to gamma-ray intensities, which then lead to an accurate quantitative determination of radioactive isotopes in the material

  5. Processing of data issued from a {gamma} spectrometer; Traitement des informations issues d'un spectrometre {gamma}

    Energy Technology Data Exchange (ETDEWEB)

    Boulanger, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1966-03-01

    The purpose of the following report is the study of computation method applied to analysis by gamma spectrometry. We first study quantitative analysis by the least-squares method, improved by a gain-shift correction. Then the single full-energy peaks and the resolution of complex peaks are dealt with. In both cases, the calculation principle, then the systematic tests achieved in order to-prove their validity and determine their application ranges are described and finally some experimental results appropriate to illustrate their possibilities are presented. (author) [French] Le present rapport a pour objet l'etude de methodes de calcul applicables a l'analyse par spectrometrie gamma. On etudie d'abord l'analyse quantitative par la methode des moindres carres, completee par une correction de derive du gain. Puis on examine les pics d'absorption totale simples et la resolution des pics complexes. Dans les deux cas, on expose le principe des calculs et les essais systematiques effectues pour eprouver leur validite et definir leurs domaines d'application, ainsi que quelques resultats experimentaux propres a illustrer leurs possibilites.

  6. Gamma histograms for radiotherapy plan evaluation

    International Nuclear Information System (INIS)

    Spezi, Emiliano; Lewis, D. Geraint

    2006-01-01

    Background and purpose: The technique known as the 'γ evaluation method' incorporates pass-fail criteria for both distance-to-agreement and dose difference analysis of 3D dose distributions and provides a numerical index (γ) as a measure of the agreement between two datasets. As the γ evaluation index is being adopted in more centres as part of treatment plan verification procedures for 2D and 3D dose maps, the development of methods capable of encapsulating the information provided by this technique is recommended. Patients and methods: In this work the concept of γ index was extended to create gamma histograms (GH) in order to provide a measure of the agreement between two datasets in two or three dimensions. Gamma area histogram (GAH) and gamma volume histogram (GVH) graphs were produced using one or more 2D γ maps generated for each slice of the irradiated volume. GHs were calculated for IMRT plans, evaluating the 3D dose distribution from a commercial treatment planning system (TPS) compared to a Monte Carlo (MC) calculation used as reference dataset. Results: The extent of local anatomical inhomogenities in the plans under consideration was strongly correlated with the level of difference between reference and evaluated calculations. GHs provided an immediate visual representation of the proportion of the treated volume that fulfilled the γ criterion and offered a concise method for comparative numerical evaluation of dose distributions. Conclusions: We have introduced the concept of GHs and investigated its applications to the evaluation and verification of IMRT plans. The gamma histogram concept set out in this paper can provide a valuable technique for quantitative comparison of dose distributions and could be applied as a tool for the quality assurance of treatment planning systems

  7. Calculating the Responses of Self-Powered Radiation Detectors.

    Science.gov (United States)

    Thornton, D. A.

    Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual

  8. Computers in activation analysis and gamma-ray spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, B. S.; D' Agostino, M. D.; Yule, H. P. [eds.

    1979-01-01

    Seventy-three papers are included under the following session headings: analytical and mathematical methods for data analysis; software systems for ..gamma..-ray and x-ray spectrometry; ..gamma..-ray spectra treatment, peak evaluation; least squares; IAEA intercomparison of methods for processing spectra; computer and calculator utilization in spectrometer systems; and applications in safeguards, fuel scanning, and environmental monitoring. Separate abstracts were prepared for 72 of those papers. (DLC)

  9. An analytical calculation of the peak efficiency for cylindrical sources perpendicular to the detector axis in gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Julio C. [Autoridad Regulatoria Nuclear, Laboratorio de Espectrometria Gamma-CTBTO, Av. Del Libertador 8250, C1429BNP Buenos Aires (Argentina)], E-mail: jaguiar@sede.arn.gov.ar

    2008-08-15

    An analytical expression for the so-called full-energy peak efficiency {epsilon}(E) for cylindrical source with perpendicular axis to an HPGe detector is derived, using point-source measurements. The formula covers different measuring distances, matrix compositions, densities and gamma-ray energies; the only assumption is that the radioactivity is homogeneously distributed within the source. The term for the photon self-attenuation is included in the calculation. Measurements were made using three different sized cylindrical sources of {sup 241}Am, {sup 57}Co, {sup 137}Cs, {sup 54}Mn, and {sup 60}Co with corresponding peaks of 59.5, 122, 662, 835, 1173, and 1332 keV, respectively, and one measurement of radioactive waste drum for 662, 1173, and 1332 keV.

  10. Lagoa Real design. Cachoeira mine. Uranium ratio from gamma profile

    International Nuclear Information System (INIS)

    Juliao, B.

    1984-06-01

    This paper presents the satisfactory accuracy of uranium ratio from gamma profile, using an equation from simple regression. The comparative study between radiometric ratios calculated from gamma data in boreholes and uranium ratio determined by Delayed Neutron Analysis shows a good measure of correlation in Cachoeira Mines. (author)

  11. Cosmic gamma-ray burst

    International Nuclear Information System (INIS)

    Yamagami, Takamasa

    1985-01-01

    Ballon experiments for searching gamma-ray burst were carried out by employing rotating-cross modulation collimators. From a very long observation of total 315 hours during 1975 to 1979, three gamma-ray intensity anomalies were observed which were speculated as a gamma-ray burst. As for the first gamma-ray intensity anomaly observed in 1975, the burst source could be located precisely but the source, heavenly body, could not be specified. Gamma-ray burst source estimation was made by analyzing distribution of burst source in the celestial sphere, burst size distribution, and burst peak. Using the above-mentioned data together with previously published ones, apparent inconsistency was found between the observed results and the adopted theory that the source was in the Galaxy, and this inconsistency was found due to the different time profiles of the burst observed with instruments of different efficiency. It was concluded by these analysis results that employment of logN - logP (relation between burst frequency and burst count) was better than that of logN - logS (burst size) in the examination of gamma-ray burst because the former was less uncertain than the latter. Analyzing the author's observed gamma-ray burst data and the related published data, it was clarified that the burst distribution was almost P -312 for the burst peak value larger than 10 -6 erg/cm 2 .sec. The author could indicate that the calculated celestial distribution of burst source was consistent with the observed results by the derivation using the logN - logP relationship and that the burst larger than 10 -6 erg/cm 2 .sec happens about one thousand times a year, about ten times of the previous value. (Takagi, S.)

  12. Evaluation of effective dose equivalent from environmental gamma rays

    International Nuclear Information System (INIS)

    Saito, K.; Tsutsumi, M.; Moriuchi, S.; Petoussi, N.; Zankl, M.; Veit, R.; Jacob, P.; Drexler, G.

    1991-01-01

    Organ doses and effective dose equivalents for environmental gamma rays were calculated using human phantoms and Monte Carlo methods accounting rigorously the environmental gamma ray fields. It was suggested that body weight is the dominant factor to determine organ doses. The weight function expressing organ doses was introduced. Using this function, the variation in organ doses due to several physical factors were investigated. A detector having gamma-ray response similar to that of human bodies has been developed using a NaI(Tl) scintillator. (author)

  13. A novel method for quantitative geosteering using azimuthal gamma-ray logging

    International Nuclear Information System (INIS)

    Yuan, Chao; Zhou, Cancan; Zhang, Feng; Hu, Song; Li, Chaoliu

    2015-01-01

    A novel method for quantitative geosteering by using azimuthal gamma-ray logging is proposed. Real-time up and bottom gamma-ray logs when a logging tool travels through a boundary surface with different relative dip angles are simulated with the Monte Carlo method. Study results show that response points of up and bottom gamma-ray logs when the logging tool moves towards a highly radioactive formation can be used to predict the relative dip angle, and then the distance from the drilling bit to the boundary surface is calculated. - Highlights: • A new method is proposed for geosteering by using azimuthal gamma-ray logging. • The new method can quantitatively determine the distance from the drilling bit to the boundary surface while the traditional geosteering method can only qualitatively guide the drilling bit in reservoirs. • The response points of real-time upper and lower gamma line when the logging tool meets high radioactive formation are used to predict the relative dip angles, and then the distance from the drilling bit to the boundary surface is calculated

  14. Determining the solar-flare photospheric scale height from SMM gamma-ray measurements

    Science.gov (United States)

    Lingenfelter, Richard E.

    1991-01-01

    A connected series of Monte Carlo programs was developed to make systematic calculations of the energy, temporal and angular dependences of the gamma-ray line and neutron emission resulting from such accelerated ion interactions. Comparing the results of these calculations with the Solar Maximum Mission/Gamma Ray Spectrometer (SMM/GRS) measurements of gamma-ray line and neutron fluxes, the total number and energy spectrum of the flare-accelerated ions trapped on magnetic loops at the Sun were determined and the angular distribution, pitch angle scattering, and mirroring of the ions on loop fields were constrained. Comparing the calculations with measurements of the time dependence of the neutron capture line emission, a determination of the He-3/H ratio in the photosphere was also made. The diagnostic capabilities of the SMM/GRS measurements were extended by developing a new technique to directly determine the effective photospheric scale height in solar flares from the neutron capture gamma-ray line measurements, and critically test current atmospheric models in the flare region.

  15. Gamma sensitivity of pressurized drift tubes

    International Nuclear Information System (INIS)

    Baranov, S.A.; Bojko, I.R.; Shelkov, G.A.; Ignatenko, M.A.

    1995-01-01

    Using a set of commonly used radioactive sources, the efficiency of pressurized drift tubes for gammas with energy from 5.9 keV up to 1.3 MeV has been measured. The tube was made of aluminium and filled with Ar, 15%CO 2 and 2.5%iC 4 H 10 gas mixture at 3 atm. The measured efficiency is compared with the results of the calculations in the frame of our simple model as well as with that of the Monte Carlo simulation using GEANT code. The results of our calculations are in agreement with experimental data, while GEANT simulation tends to give lower efficiency in the energy range of 200 keV γ <1300 keV. The average efficiency of the tube in the field of ATLAS gamma background is about 0.45%. 8 refs., 7 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  17. Monitoring gamma radioactivity over large land areas using portable equipment

    International Nuclear Information System (INIS)

    Mac Mahon, T.D.; Gray, P.W.; Eer, A.M. D'; Naboulsi, A.H.; Koutsoyannopoulos, C.

    1990-01-01

    The principal objective of this research has been to provide information on cost-effective techniques to detect localized areas of gamma-emitting radionuclides. This objective has been achieved by determining the time required to scan unit area as a function of depth of the gamma source below the site surface, the activity of the gamma source, the energy of the emitted gamma-ray, and the gamma transport properties of the site material. A comparison between survey and sampling techniques is made, and the advantages of using survey techniques to detect localized gamma-ray sources are discussed. A survey technique based on an adaptive moving array detector system is described. A field experiment has been carried out to verify the results of calculations of the sensitivity of the techniques described

  18. Photoproduction data for heating calculations

    International Nuclear Information System (INIS)

    Van der Marck, Steven C.; Koning, Arjan J.; Rochman, Dimitri

    2008-01-01

    For irradiations in a materials test reactor, the prediction of the amount of gamma heating in the reactor is important. Only a good predictive calculation will lead to an irradiation in which the specified temperatures are reached. The photons produced by fission product decay are often missing in spectrum calculations for a reactor, but the contribution of the photons can be computed effectively using engineering correlations for the amount of fission product decay and the ensuing photon spectrum. The prompt photons are usually calculated by a spectrum code based on the underlying nuclear data libraries. For most of the important nuclides, the nuclear data libraries contain data for the photon productions rates. However, there are still many nuclides for which the photon production data are missing, and some of these nuclides contribute to gamma heating. In this paper it is estimated what the contributions to heating are from photon production on nuclides such as 236 U, 238 Pu, 135 I, 135 Xe, 147 Pm, 148 Pm, 148m Pm, and 149 Sm. Also, simple arguments are given to judge the effect from photon production on all other (lumped) fission products, and from 28 Al decay. For all these calculations the High Flux Reactor is used as an example. (authors)

  19. Three-dimensional neutron dose distribution in the environment around a 1-GeV electron synchrotron facility at INS

    International Nuclear Information System (INIS)

    Uwamino, Y.; Nakamura, T.

    1987-01-01

    The three-dimensional (surface and altitude) skyshine neutron-dose-equivalent distribution around the 1-GeV electron synchrotron (ES) of the Institute for Nuclear Study, University of Tokyo, was measured with a high-sensitivity dose-equivalent counter. The neutron spectrum in the environment was also measured with a multimoderator spectrometer incorporating a 3 He counter. The dose-equivalent distribution and the leakage neutron spectrum at the surface of the ES building were measured with a Studsvik 2202D counter and the multimoderator spectrometer, including an indium activation detector. Skyshine neutron transport calculations, beginning with the photoneutron spectrum and yielding the dose-equivalent distribution in the environment, were performed with the DOT3.5 code and two Monte Carlo codes, MMCR-2 and MMCR-3, using the DLC-87/HILO group cross sections. The calculated neutron spectra at the top surface of the concrete ceiling and at a point 111 m from the ES agreed well with the measured results, and the calculated three-dimensional dose-equivalent distribution also agreed. The dose value increased linearly with altitude, and the slope was estimated for neutron-producing facilities. (author)

  20. Monte-Carlo method applied to the energy loss calculation of the gamma rays isotropic flux in the NaI(tau l) cylindrical scintillator between 0.5-20 MeV

    International Nuclear Information System (INIS)

    Martin, I.M.; Dutra, S.L.G.; Palmeira, R.A.R.

    1975-01-01

    Using the 'Monte Carlo' method, a determination was made of the response function of a NaI cylindrical crystal when exposed to an omnidirectional γ ray flux in the range 0.5 - 20 MeV. Improvements over previous similar calculations include considerations of the bremsstrahlung and multiple scattering processes in the slowing down of the secondary electrons. These calculations will be applied to the problem of determining the energy spectrum of an incident gamma ray flux from the measured response of the crystal in the space [pt

  1. Stieltjes-moment-theory technique for calculating resonance width's

    International Nuclear Information System (INIS)

    Hazi, A.U.

    1978-12-01

    A recently developed method for calculating the widths of atomic and molecular resonances is reviewed. The method is based on the golden-rule definition of the resonance width, GAMMA(E). The method uses only square-integrable, L 2 , basis functions to describe both the resonant and the non-resonant parts of the scattering wave function. It employs Stieltjes-moment-theory techniques to extract a continuous approximation for the width discrete representation of the background continuum. Its implementation requires only existing atomic and molecular structure codes. Many-electron effects, such as correlation and polarization, are easily incorporated into the calculation of the width via configuration interaction techniques. Once the width, GAMMA(E), has been determined, the energy shift can be computed by a straightforward evaluation of the required principal-value integral. The main disadvantage of the method is that it provides only the total width of a resonance which decays into more than one channel in a multichannel problem. A review of the various aspects of the theory is given first, and then representative results that have been obtained with this method for several atomic and molecular resonances are discussed. 28 references, 3 figures, 4 tables

  2. Calculating radiation exposure and dose

    International Nuclear Information System (INIS)

    Hondros, J.

    1987-01-01

    This paper discusses the methods and procedures used to calculate the radiation exposures and radiation doses to designated employees of the Olympic Dam Project. Each of the three major exposure pathways are examined. These are: gamma irradiation, radon daughter inhalation and radioactive dust inhalation. A further section presents ICRP methodology for combining individual pathway exposures to give a total dose figure. Computer programs used for calculations and data storage are also presented briefly

  3. PSYCRODATA: a software which calculates the air humidity characteristics and relate its with the variations of the gamma environmental bottom

    International Nuclear Information System (INIS)

    Alonso A, D.; Dominguez L, O.; Ramos V, O.; Caveda R, C.A.; Capote F, E.; Dominguez G, A.; Valdes S, E.; Rodriguez V, E.

    2006-01-01

    The computer tool 'Psycrodata', able to calculate the values of those characteristics of the humidity of the air starting from the measurements carried out of humidity and temperature in the post of occident of the National Net of Environmental Radiological Surveillance was obtained. Among the facilities that 'Psycrodata' toasts it is the keeping the obtained information in a database facilitating the making of reports. For another part the possibility of selection of different approaches for the calculation and the introduction of the psicrometric coefficient to use, its make that each station can have the suitable psicrometric chart keeping in mind the instrumentation and the characteristics of the area of location of the same one. Also, can have facilities to import text files for later on to be plotted, it allowed to correlate the absorbed dose rate in air due to the environmental gamma radiation, besides of the temperature and the humidity, with the tension of the water steam, the temperature of the dew point and the saturation deficit. (Author)

  4. The Effect of Material Homogenization in Calculating the Gamma-Ray dose from Spent PWR Fuel Pins in an Air Medium

    International Nuclear Information System (INIS)

    TH Trumbull

    2005-01-01

    The effect of material homogenization on the calculated dose rate was studied for several arrangements of typical PWR spent fuel pins in an air medium using the Monte Carlo code, MCNP. The models analyzed increased in geometric complexity, beginning with a single fuel pin, progressing to ''small'' lattices, i.e., 3x3, 5x5, 7x7 fuel pins, and culminating with a full 17x17 pin PWR bundle analysis. The fuel pin dimensions and compositions were taken directly from a previous study and efforts were made to parallel this study by specifying identical flux-to-dose functions and gamma-ray source spectra. The analysis shows two competing components to the overall effect of material homogenization on calculated dose rate. Homogenization of pin lattices tends to lower the effect of radiation ''channeling'' but increase the effect of ''source redistribution.'' Depending on the size of the lattice and location of the detectors, the net effect of material homogenization on dose rate can be insignificant or range from a 6% decrease to a 35% increase relative to the detailed geometry model

  5. $\\gamma$-$\\gamma$ and $\\gamma$-p events at high energies

    CERN Document Server

    Schuler, Gerhard A.; Gerhard A Schuler; Torbjorn Sjostrand

    1994-01-01

    A real photon has a complicated nature, whereby it may remain unresolved or fluctuate into a vector meson or a perturbative q-qbar pair. Based on this picture, we previously presented a model for gamma-p events that is based on the presence of three main event classes: direct, VMD and anomalous. In gamma-gamma events, a natural generalization gives three-by-three combinations of the nature of the two incoming photons, and thus six distinct event classes. The properties of these classes are constrained by the choices already made, in the gamma-p model, of cut-off procedures and other aspects. It is therefore possible to predict the energy-dependence of the cross section for each of the six components separately. The total cross section thus obtained is in good agreement with data, and also gives support to the idea that a simple factorized ansatz with a pomeron and a reggeon term can be a good approximation. Event properties undergo a logical evolution from p-p to gamma-p to gamma-gamma events, with larger cha...

  6. JADSPE, Multi-Channel Gamma Spectra Unfolding Program

    International Nuclear Information System (INIS)

    Rikovska, J.; Stejskalova, E.

    2005-01-01

    1 - Description of program or function: JADSPE is a package of eight programs to process multi-channel gamma-ray spectra. The programs can be used to: - locate automatically spectral peaks and calculate their positions, areas, and full widths at half maximum (FWHM); - plot the spectra on a CALCOMP plotter, TEKTRONIX terminal or a line printer; - add or subtract several spectra with the possibility of adjusting either their start and end channels or the maxima of the chosen corresponding peaks. The JADSPE package comprises the following programs: - SPECTF: automatic location of peaks and calculation of their positions, areas and FWHMS. The standard deviations of peak parameters are also determined, and each evaluated region is plotted on the line printer. - SPECT1: The areas and FWHMs are calculated for peaks whose positions are known beforehand. The standard deviations of calculated parameters are also determined, and each evaluated region is plotted on the line printer. - PLOCHA: The peak net area is calculated by summing the channel contents in specified regions and by subtracting a linear background. - GRAPH: Spectrum plotting on the line printer. - PLTNEW: Spectrum plotting on CALCOMP plotter or on TEKTRONIX terminal. - SUMDIF: The channel contents of several gamma-ray spectra are added or subtracted. - SSPFP: The channel contents of several gamma-ray spectra are added with adjustment of the maxima of specified peaks. - SOUCET: The channel contents of several gamma-ray spectra are added with the adjustment of start and end channels of the spectra. 2 - Method of solution: Non-linear least-square fit. 3 - Restrictions on the complexity of the problem: The full energy peaks are approximated by a symmetrical Gaussian function and the underlying background is approximated by a first-order polynomial. A fixed spectrum length of 4096 channels is assumed. Maxima of: - number of peaks in one multiplet: 9; - number of peaks identified by the automatic search procedure

  7. Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Nilsson, R

    1964-09-15

    Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B{sub 3}), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources.

  8. Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

    International Nuclear Information System (INIS)

    Aalto, E.; Nilsson, R.

    1964-09-01

    Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B 3 ), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources

  9. /sup 56/Fe (. gamma. ,. cap alpha. /sub 0/) reaction

    Energy Technology Data Exchange (ETDEWEB)

    Tamae, T; Sugawara, M [Tohoku Univ., Sendai (Japan). Lab. of Nuclear Science; Tsubota, H

    1974-12-01

    The reaction cross section of /sup 56/Fe (..gamma.., ..cap alpha../sub 0/) was measured from the electron energy of 15 to 25 MeV. The measured data were compared with the calculated ones based on statistic theory. Both agreed with each other. Therefore, the affirmative result was obtained for the presumption that the reaction of (..gamma.., ..cap alpha../sub 0/) of the nuclei around these energy levels can be explained by the statistical theory. The angular distribution of /sup 56/Fe (..gamma.., ..cap alpha../sub 0/) with 17 MeV electron energy was also measured, and the E2/E1 ratio was obtained. In the measurement of the /sup 56/Fe ( Gamma , ..cap alpha../sub 0/) reaction cross section, a natural target of 2.69 mg/cm/sup 2/ was irradiated with an electron beam with energy from 15 MeV to 25 MeV at intervals of 0.5 MeV, and the emitted ..cap alpha.. particles were detected by a broad band magnetic distribution meter. The measured cross section of the (..gamma.., ..cap alpha../sub 0/) reaction agreed with the calculated one based on statistical theory. If this fact is recognized in many nuclei, the cross section of the (..gamma.., ..cap alpha../sub 0/) reaction on those nuclei has the following characteristics. When the increasing rate of the product of a complex nucleus formation cross section and ..cap alpha../sub 0/ penetration factor is larger than that of the sum of all penetration factors of possible channels, the cross section of the (..gamma.., ..cap alpha../sub 0/) reaction increases, and takes a peak value when the above two increasing rates agree with each other.

  10. Displacement damage caused by gamma-rays and neutrons on Au and Se.

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, Barney Lee [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-11-01

    This report documents theoretical calculations of displacement damage produced by gamma rays and neutrons on various materials. The average energy of the gamma rays was 1.24 MeV and 1.0 MeV for the neutrons. The fluence of the gamma rays was 1.2e14 γ/cm2 , for the neutrons it was 1.0e12 n/cm2. The initial materials of interest were Au and Se. The total doses of the gamma ray exposures were in the 100 kRad range for both elements. An equivalent electron fluence was approximated to be the same as the gamma ray fluence over one gamma ray attenuation length in both materials and at the same 1.24 MeV energy. The maximum recoil energy of the Au and Se for these electrons was calculated relativisticaly to be 29 and 72 eV respectively. The relativisitic McKinley and Feshbach theory for the atomic recoil cross sections produced by the electrons were in the 10s of mbarn range and an upper limit for the concentration of Frenkel pairs for the gamma ray exposures for both elements was in the ppb range. The Robinson Energy Partioning Theory for non-ionizing energy loss (NIEL) of ions in solids was used to calculate the concentration of Frenkel pairs produced by the 1 MeV neutrons, and this concentration was also in the ppb range for both Au and Se. Low damage levels like this can have effects on minority carrier recombination in semiconductors, but are not expected to have any effect on metals like Au, or metalloids such as Se.

  11. Research building gamma Compton scattering measurement system and related exercises for training nuclear human resources

    International Nuclear Information System (INIS)

    Mai Xuan Phong; Nguyen Van Hung; Pham Xuan Hai; Le Van Ngoc; Nguyen Xuan Hai; Dang Lanh; Tran Quoc Duong

    2013-01-01

    In this subject we have designed and manufactured Compton scattering gamma measurement system based on the calculated optimal configuration as well as the conditions of protect radiation by using Monte-Carlo simulation program and fabrication with the optimal conditions were selected. Monte-Carlo simulation calculation of Compton scattering gamma follow different angles on copper, surveying gamma radiation attenuation characteristics of materials: lead, iron, aluminum, and compared with the experimental results performed on the same measurement system has been built and given for evaluation, comments. (author)

  12. Precise measurement of {gamma}(K{yields}e {nu}({gamma}))/{gamma}(K{yields}{mu} {nu}({gamma})) and study of K{yields}e {nu} {gamma}

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosino, F.; Massarotti, P.; Meola, S.; Napolitano, M. [Dipartimento di Scienze Fisiche dell' Universita ' ' Federico II' ' , Napoli (Italy); INFN Sezione di Napoli, Napoli (Italy); Antonelli, A.; Antonelli, M.; Bencivenni, G.; Bloise, C.; Bossi, F.; Capon, G.; Capussela, T.; Ciambrone, P.; De Lucia, E.; De Simone, P.; Dreucci, M.; Felici, G.; Gatti, C.; Giovannella, S.; Jacewicz, M.; Lanfranchi, G.; Miscetti, S.; Moulson, M.; Murtas, F.; Palutan, M.; Santangelo, P.; Sciascia, B.; Sibidanov, A.; Spadaro, T.; Venanzoni, G. [Laboratori Nazionali di Frascati dell' INFN, Frascati (Italy); Archilli, F. [Dipartimento di Fisica dell' Universita ' ' Tor Vergata' ' , Rome (Italy); INFN Sezione di Roma Tor Vergata, Rome (Italy); Beltrame, P.; Denig, A.; Mueller, S. [Johannes Gutenberg-Universitaet, Institut fuer Kernphysik, Mainz (Germany); Bini, C.; De Santis, A.; De Zorzi, G.; Di Domenico, A.; Fiore, S.; Franzini, P.; Gauzzi, P. [Dipartimento di Fisica dell' Universita ' ' La Sapienza' ' , Rome (Italy); INFN Sezione di Roma, Rome (Italy); Bocchetta, S.; Ceradini, F.; Di Micco, B.; Nguyen, F. [Dipartimento di Fisica dell' Universita ' ' Roma Tre' ' , Rome (Italy); INFN Sezione di Roma Tre, Rome (Italy); Branchini, P.; Graziani, E.; Passeri, A.; Tortora, L. [INFN Sezione di Roma Tre, Rome (Italy); Capriotti, D. [Dipartimento di Fisica dell' Universita ' ' Roma Tre' ' , Rome (Italy); Di Donato, C. [INFN Sezione di Napoli, Napoli (Italy); Kulikov, V. [Institute for Theoretical and Experimental Physics, Moscow (Russian Federation); Lee-Franzini, J. [Laboratori Nazionali di Frascati dell' INFN, Frascati (Italy); State University of New York, Physics Department, Stony Brook (United States); Martini, M.; Patera, V.; Versaci, R. [Laboratori Nazionali di Frascati dell' INFN, Frascati (Italy); Dipartimento di Energetica dell' Universita ' ' La Sapienza' ' , Rome (Italy); Valente, P. [INFN Sezione di Roma, Rome (Italy)

    2009-12-15

    We present a precise measurement of the ratio R{sub K}={gamma}(K{yields}e{nu}({gamma}))/{gamma}(K{yields}{mu}{nu}({gamma})) and a study of the radiative process K{yields}e{nu}{gamma}, performed with the KLOE detector. The results are based on data collected at the Frascati e{sup +}e{sup -} collider DA {phi}NE for an integrated luminosity of 2.2 fb{sup -1}. We find R{sub K}=(2.493{+-}0.025{sub stat}{+-}0.019{sub syst}) x 10{sup -5}, in agreement with the Standard Model expectation. This result is used to improve constraints on parameters of the Minimal Supersymmetric Standard Model with lepton flavor violation. We also measured the differential decay rate d {gamma}(K{yields}e{nu}{gamma})/dE{sub {gamma}} for photon energies 10gamma}}<250 MeV. Results are compared with predictions from theory. (orig.)

  13. Measurements of the Cross Sections for Open Charm and Beauty Production in $\\gamma\\gamma$ Collisions at $\\sqrt{s}$ = 189-202 GeV

    CERN Document Server

    Acciarri, M.; Adriani, O.; Aguilar-Benitez, M.; Alcaraz, J.; Alemanni, G.; Allaby, J.; Aloisio, A.; Alviggi, M.G.; Ambrosi, G.; Anderhub, H.; Andreev, Valery P.; Angelescu, T.; Anselmo, F.; Arefev, A.; Azemoon, T.; Aziz, T.; Bagnaia, P.; Bajo, A.; Baksay, L.; Balandras, A.; Baldew, S.V.; Banerjee, S.; Banerjee, Sw.; Barczyk, A.; Barillere, R.; Bartalini, P.; Basile, M.; Batalova, N.; Battiston, R.; Bay, A.; Becattini, F.; Becker, U.; Behner, F.; Bellucci, L.; Berbeco, R.; Berdugo, J.; Berges, P.; Bertucci, B.; Betev, B.L.; Bhattacharya, S.; Biasini, M.; Biland, A.; Blaising, J.J.; Blyth, S.C.; Bobbink, G.J.; Bohm, A.; Boldizsar, L.; Borgia, B.; Bourilkov, D.; Bourquin, M.; Braccini, S.; Branson, J.G.; Brochu, F.; Buffini, A.; Buijs, A.; Burger, J.D.; Burger, W.J.; Cai, X.D.; Capell, M.; Cara Romeo, G.; Carlino, G.; Cartacci, A.M.; Casaus, J.; Castellini, G.; Cavallari, F.; Cavallo, N.; Cecchi, C.; Cerrada, M.; Cesaroni, F.; Chamizo, M.; Chang, Y.H.; Chaturvedi, U.K.; Chemarin, M.; Chen, A.; Chen, G.; Chen, G.M.; Chen, H.F.; Chen, H.S.; Chiefari, G.; Cifarelli, L.; Cindolo, F.; Civinini, C.; Clare, I.; Clare, R.; Coignet, G.; Colino, N.; Costantini, S.; Cotorobai, F.; de la Cruz, B.; Csilling, A.; Cucciarelli, S.; Dai, T.S.; van Dalen, J.A.; D'Alessandro, R.; de Asmundis, R.; Deglon, P.; Degre, A.; Deiters, K.; della Volpe, D.; Delmeire, E.; Denes, P.; DeNotaristefani, F.; De Salvo, A.; Diemoz, M.; Dierckxsens, M.; van Dierendonck, D.; Dionisi, C.; Dittmar, M.; Dominguez, A.; Doria, A.; Dova, M.T.; Duchesneau, D.; Dufournaud, D.; Duinker, P.; Duran, I.; El Mamouni, H.; Engler, A.; Eppling, F.J.; Erne, F.C.; Ewers, A.; Extermann, P.; Fabre, M.; Falagan, M.A.; Falciano, S.; Favara, A.; Fay, J.; Fedin, O.; Felcini, M.; Ferguson, T.; Fesefeldt, H.; Fiandrini, E.; Field, J.H.; Filthaut, F.; Fisher, P.H.; Fisk, I.; Forconi, G.; Freudenreich, K.; Furetta, C.; Galaktionov, Iouri; Ganguli, S.N.; Garcia-Abia, Pablo; Gataullin, M.; Gau, S.S.; Gentile, S.; Gheordanescu, N.; Giagu, S.; Gong, Z.F.; Grenier, Gerald Jean; Grimm, O.; Gruenewald, M.W.; Guida, M.; van Gulik, R.; Gupta, V.K.; Gurtu, A.; Gutay, L.J.; Haas, D.; Hasan, A.; Hatzifotiadou, D.; Hebbeker, T.; Herve, Alain; Hidas, P.; Hirschfelder, J.; Hofer, H.; Holzner, G.; Hoorani, H.; Hou, S.R.; Hu, Y.; Iashvili, I.; Jin, B.N.; Jones, Lawrence W.; de Jong, P.; Josa-Mutuberria, I.; Khan, R.A.; Kafer, D.; Kaur, M.; Kienzle-Focacci, M.N.; Kim, D.; Kim, J.K.; Kirkby, Jasper; Kiss, D.; Kittel, W.; Klimentov, A.; Konig, A.C.; Kopal, M.; Kopp, A.; Koutsenko, V.; Kraber, M.; Kraemer, R.W.; Krenz, W.; Kruger, A.; Kunin, A.; Ladron de Guevara, P.; Laktineh, I.; Landi, G.; Lebeau, M.; Lebedev, A.; Lebrun, P.; Lecomte, P.; Lecoq, P.; Le Coultre, P.; Lee, H.J.; Le Goff, J.M.; Leiste, R.; Levtchenko, P.; Li, C.; Likhoded, S.; Lin, C.H.; Lin, W.T.; Linde, F.L.; Lista, L.; Liu, Z.A.; Lohmann, W.; Longo, E.; Lu, Y.S.; Lubelsmeyer, K.; Luci, C.; Luckey, David; Lugnier, L.; Luminari, L.; Lustermann, W.; Ma, W.G.; Maity, M.; Malgeri, L.; Malinin, A.; Mana, C.; Mangeol, D.; Mans, J.; Marian, G.; Martin, J.P.; Marzano, F.; Mazumdar, K.; McNeil, R.R.; Mele, S.; Merola, L.; Meschini, M.; Metzger, W.J.; von der Mey, M.; Mihul, A.; Milcent, H.; Mirabelli, G.; Mnich, J.; Mohanty, G.B.; Moulik, T.; Muanza, G.S.; Muijs, A.J.M.; Musicar, B.; Musy, M.; Napolitano, M.; Nessi-Tedaldi, F.; Newman, H.; Niessen, T.; Nisati, A.; Kluge, Hannelies; Ofierzynski, R.; Organtini, G.; Oulianov, A.; Palomares, C.; Pandoulas, D.; Paoletti, S.; Paolucci, P.; Paramatti, R.; Park, H.K.; Park, I.H.; Passaleva, G.; Patricelli, S.; Paul, Thomas Cantzon; Pauluzzi, M.; Paus, C.; Pauss, F.; Pedace, M.; Pensotti, S.; Perret-Gallix, D.; Petersen, B.; Piccolo, D.; Pierella, F.; Pieri, M.; Piroue, P.A.; Pistolesi, E.; Plyaskin, V.; Pohl, M.; Pojidaev, V.; Postema, H.; Pothier, J.; Prokofev, D.O.; Prokofiev, D.; Quartieri, J.; Rahal-Callot, G.; Rahaman, M.A.; Raics, P.; Raja, N.; Ramelli, R.; Rancoita, P.G.; Ranieri, R.; Raspereza, A.; Raven, G.; Razis, P.; Ren, D.; Rescigno, M.; Reucroft, S.; Riemann, S.; Riles, Keith; Rodin, J.; Roe, B.P.; Romero, L.; Rosca, A.; Rosier-Lees, S.; Roth, Stefan; Rosenbleck, C.; Rubio, J.A.; Ruggiero, G.; Rykaczewski, H.; Saremi, S.; Sarkar, S.; Salicio, J.; Sanchez, E.; Sanders, M.P.; Schafer, C.; Schegelsky, V.; Schmidt-Kaerst, S.; Schmitz, D.; Schopper, H.; Schotanus, D.J.; Schwering, G.; Sciacca, C.; Seganti, A.; Servoli, L.; Shevchenko, S.; Shivarov, N.; Shoutko, V.; Shumilov, E.; Shvorob, A.; Siedenburg, T.; Son, D.; Smith, B.; Spillantini, P.; Steuer, M.; Stickland, D.P.; Stone, A.; Stoyanov, B.; Straessner, A.; Sudhakar, K.; Sultanov, G.; Sun, L.Z.; Sushkov, S.; Suter, H.; Swain, J.D.; Szillasi, Z.; Sztaricskai, T.; Tang, X.W.; Tauscher, L.; Taylor, L.; Tellili, B.; Timmermans, Charles; Ting, Samuel C.C.; Ting, S.M.; Tonwar, S.C.; Toth, J.; Tully, C.; Tung, K.L.; Uchida, Y.; Ulbricht, J.; Valente, E.; Vesztergombi, G.; Vetlitsky, I.; Vicinanza, D.; Viertel, G.; Villa, S.; Vivargent, M.; Vlachos, S.; Vodopianov, I.; Vogel, H.; Vogt, H.; Vorobev, I.; Vorobov, A.A.; Vorvolakos, A.; Wadhwa, M.; Wallraff, W.; Wang, M.; Wang, X.L.; Wang, Z.M.; Weber, A.; Weber, M.; Wienemann, P.; Wilkens, H.; Wu, S.X.; Wynhoff, S.; Xia, L.; Xu, Z.Z.; Yamamoto, J.; Yang, B.Z.; Yang, C.G.; Yang, H.J.; Yang, M.; Ye, J.B.; Yeh, S.C.; Zalite, An.; Zalite, Yu.; Zhang, Z.P.; Zhu, G.Y.; Zhu, R.Y.; Zichichi, A.; Zilizi, G.; Zimmermann, B.; Zoller, M.

    2001-01-01

    The production of c and b quarks in gamma-gamma collisions is studied with the L3 detector at LEP with 410 pb^-1 of data, collected at centre-of-mass energies from 189 GeV to 202 GeV. Hadronic final states containing c and b quarks are identified by detecting electrons or muons from their semileptonic decays. The cross sections sigma(e+e- -> e+e- c c~ X) and sigma(e+e- -> e+e- b b~ X) are measured and compared to next-to-leading order perturbative QCD calculations. The cross section of b production is measured in gamma-gamma collisions for the first time. It is in excess of the QCD prediction by a factor of three.

  14. Gamma-ray emission spectra from spheres with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

    1989-01-01

    Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

  15. Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

    International Nuclear Information System (INIS)

    Quade, U.

    1994-01-01

    Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

  16. In situ measurements of dose rates from terrestrial gamma rays

    International Nuclear Information System (INIS)

    Horng, M.C.; Jiang, S.H.

    2002-01-01

    A portable, high purity germanium (HPGe) detector was employed for the performance of in situ measurements of radionuclide activity concentrations in the ground in Taiwan, at altitudes ranging from sea level to 3900 m. The absolute peak efficiency of the HPGe detector for a gamma-ray source uniformly distributed in the semi-infinite ground was determined using a semi-empirical method. The gamma-ray dose rates from terrestrial radionuclides were calculated from the measured activity levels using recently published dose rate conversion factors. The absorbed dose rate in air due to cosmic rays was derived by subtracting the terrestrial gamma-ray dose rate from the overall absorbed dose rate in air measured using a high-pressure ionization chamber. The cosmic-ray dose rate calculated as a function of altitude, was found to be in good agreement with the data reported by UNSCEAR. (orig.)

  17. Prompt gamma-ray analysis of steel slag in concrete

    International Nuclear Information System (INIS)

    Naqvi, Akhtar Abbas; Garwan, Muhammad Ahmad; Nagadi, Mahmoud Mohammad; Rehman, Khateeb-ur; Raashid, Mohammad; Masalehuddin Mohiuddin, Mohammad; Al-Amoudi, Omar Saeed Baghabra

    2009-01-01

    Blast furnace slag (BFS) is added to Portland cement concrete to increase its durability, particularly its corrosion resistance. Monitoring the concentration of BFS in concrete for quality control purposes is desired. In this study, the concentration of BFS in concrete was measured by utilizing an accelerator-based prompt gamma-ray neutron activation analysis (PGNAA) setup. The optimum size of the BFS cement concrete specimen that produces the maximum intensity of gamma rays at the detector location was calculated through Monte Carlo simulations. The simulation results were experimentally validated through the gamma-ray yield measurement from BFS cement concrete specimens having different radii. The concentration of BFS in the cement concrete specimens was assessed through calcium and silicon gamma-ray yield measurement from cement concrete specimens containing 5 to 80 wt% BFS. The yield of calcium gamma rays decreases with increasing BFS concentration in concrete while the yield of silicon gamma rays increases with increasing BFS concentration in concrete. The calcium-to-silicon gamma-ray yield ratio has an inverse relation with BFS concentration in concrete. (author)

  18. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  19. Gamma-ray pulsars: Emission zones and viewing geometries

    Science.gov (United States)

    Romani, Roger W.; Yadigaroglu, I.-A.

    1995-01-01

    There are now a half-dozen young pulsars detected in high-energy photons by the Compton Gamma-Ray Observatory (CGRO), showing a variety of emission efficiencies and pulse profiles. We present here a calculation of the pattern of high-energy emission on the sky in a model which posits gamma-ray production by charge-depleted gaps in the outer magnetosphere. This model accounts for the radio to gamma-ray pulse offsets of the known pulsars, as well as the shape of the high-energy pulse profiles. We also show that about one-third of emitting young radio pulsars will not be detected due to beaming effects, while approximately 2.5 times the number of radio-selected gamma-ray pulsars will be viewed only high energies. Finally we compute the polarization angle variation and find that the previously misunderstood optical polarization sweep of the Crab pulsar arises naturally in this picture. These results strongly support an outer magnetosphere location for the gamma-ray emission.

  20. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  1. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  2. The pilot plant in Geiselbullach for the gamma irradiation of sewage sludge - design, operation experience and cost calculations

    International Nuclear Information System (INIS)

    Lessel, T.; Hennig, E.

    1976-01-01

    Gamma irradiation of sewage sludge is possible with facilities of simple design and great availability; they can be working fully automatically 24 hours on 350 days a year or more. No specially trained service staff is necessary. The costs for gamma irradiation of sewage sludge are slightly higher than for heat treatment, but several secondary effects speak in favour of the irradiated sludge. The hygienization of sewage sludge by gamma irradiation is normally only useful when sludge has to be disinfected during the whole year. (orig.) [de

  3. Calculation of direct effects of {sup 60}Co gamma rays on the different DNA structural levels: A simulation study using the Geant4-DNA toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Tajik, Marjan; Rozatian, Amir S.H. [Department of Physics, University of Isfahan, Hezar Jarib Street, Isfahan 81746-73441 (Iran, Islamic Republic of); Semsarha, Farid, E-mail: Semsarha@ibb.ut.ac.ir [Institute of Biochemistry and Biophysics (IBB), University of Tehran, P.O. Box: 13145-1384, Tehran (Iran, Islamic Republic of)

    2015-03-01

    In this study, simple single strand breaks (SSB) and double strand breaks (DSB) due to direct effects of the secondary electron spectrum of {sup 60}Co gamma rays on different organizational levels of a volume model of the B-DNA conformation have been calculated using the Geant4-DNA toolkit. Result of this study for the direct DSB yield shows a good agreement with other theoretical and experimental results obtained by both photons and their secondary electrons; however, in the case of SSB a noticeable difference can be observed. Moreover, regarding the almost constant yields of the direct strand breaks in the different structural levels of the DNA, calculated in this work, and compared with some theoretical studies, it can be deduced that the direct strand breaks yields depend mainly on the primary double helix structure of the DNA and the higher-order structures cannot have a noticeable effect on the direct DNA damage inductions by {sup 60}Co gamma rays. In contrast, a direct dependency between the direct SSB and DSB yields and the volume of the DNA structure has been found. Also, a further study on the histone proteins showed that they can play an important role in the trapping of low energy electrons without any significant effect on the direct DNA strand breaks inductions, at least in the range of energies used in the current study.

  4. AIRGAMMA, External Gamma-Ray Exposure from Radioactive Cloud

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Iijima, Tshinori

    1989-01-01

    1 - Description of program or function: AIRGAMMA calculates quickly the external exposure to gamma rays from a radioactive cloud. 2 - Method of solution: The external exposure is calculated by interpolating the normalized doses providing on the basis of the Gaussian plume model. 3 - Restrictions on the complexity of the problem: Memory requirement is 30 Kbytes

  5. IMRT QA: Selecting gamma criteria based on error detection sensitivity

    Energy Technology Data Exchange (ETDEWEB)

    Steers, Jennifer M. [Department of Radiation Oncology, Cedars-Sinai Medical Center, Los Angeles, California 90048 and Physics and Biology in Medicine IDP, David Geffen School of Medicine, University of California, Los Angeles, Los Angeles, California 90095 (United States); Fraass, Benedick A., E-mail: benedick.fraass@cshs.org [Department of Radiation Oncology, Cedars-Sinai Medical Center, Los Angeles, California 90048 (United States)

    2016-04-15

    Purpose: The gamma comparison is widely used to evaluate the agreement between measurements and treatment planning system calculations in patient-specific intensity modulated radiation therapy (IMRT) quality assurance (QA). However, recent publications have raised concerns about the lack of sensitivity when employing commonly used gamma criteria. Understanding the actual sensitivity of a wide range of different gamma criteria may allow the definition of more meaningful gamma criteria and tolerance limits in IMRT QA. We present a method that allows the quantitative determination of gamma criteria sensitivity to induced errors which can be applied to any unique combination of device, delivery technique, and software utilized in a specific clinic. Methods: A total of 21 DMLC IMRT QA measurements (ArcCHECK®, Sun Nuclear) were compared to QA plan calculations with induced errors. Three scenarios were studied: MU errors, multi-leaf collimator (MLC) errors, and the sensitivity of the gamma comparison to changes in penumbra width. Gamma comparisons were performed between measurements and error-induced calculations using a wide range of gamma criteria, resulting in a total of over 20 000 gamma comparisons. Gamma passing rates for each error class and case were graphed against error magnitude to create error curves in order to represent the range of missed errors in routine IMRT QA using 36 different gamma criteria. Results: This study demonstrates that systematic errors and case-specific errors can be detected by the error curve analysis. Depending on the location of the error curve peak (e.g., not centered about zero), 3%/3 mm threshold = 10% at 90% pixels passing may miss errors as large as 15% MU errors and ±1 cm random MLC errors for some cases. As the dose threshold parameter was increased for a given %Diff/distance-to-agreement (DTA) setting, error sensitivity was increased by up to a factor of two for select cases. This increased sensitivity with increasing dose

  6. Gamma ray and neutron shielding properties of some concrete materials

    International Nuclear Information System (INIS)

    Yilmaz, E.; Baltas, H.; Kiris, E.; Ustabas, I.; Cevik, U.; El-Khayatt, A.M.

    2011-01-01

    Highlights: → This study sheds light on the shielding properties of gamma-rays and neutrons for some concrete samples. → The experimental mass attenuation coefficients values were compared with theoretical values obtained using WinXCom. → Moreover, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. → The NXcom program was employed to calculate the attenuation coefficients values of neutrons. → These values showed a change with energy and composition of the concrete samples. - Abstract: Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.

  7. Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

    International Nuclear Information System (INIS)

    Ardani

    2011-01-01

    Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

  8. Development of aerial gamma radiation survey system, 4

    International Nuclear Information System (INIS)

    Saito, Kimiaki; Nagaoka, Toshi; Sakamoto, Ryuichi; Tsutsumi, Masahiro; Moriuchi, Shigeru

    1985-02-01

    Field experiments have been performed by JAERI since 1982 to obtain fundamental data required for development of aerial radiation survey system. In order to supplement the fundamental radiation data, theoretical calculations have been carried out. The utilized Monte Carlo transport program was verified by simulative calculations of the field experiments, and characteristics data on environmental gamma rays have been accumulated. In this report, the field experiments in 1981 and 1982 were simulated making use of the Monte Carlo transport calculation code YURI developed in JAERI. Comparisons were made between experimental and calculated results for exposure rate and flux density originated from terrestrial sources, and from a point source at height of 2.5 m above the ground. Good agreements between the data verified the transport program. As fundamental characteristics data on environmental gamma rays, spatial distributions of exposure, fluence, energy spectra, angular spectra and average energy were reported and discussed, for terrestrial sources of 40 K, 232 Th-series and 238 U-series, for a plane source on the ground and for a point source at 2.5 m above the ground. (author)

  9. Initial measurement of site boundary neutron dose and comparison with calculations

    International Nuclear Information System (INIS)

    Degtyarenko, P.; Dotson, D.; May, R.; Schwahn, S.; Stapleton, G.

    1996-01-01

    For most accelerators adequate side shielding can be provided at minimal cost to meet the most aggressive radiation protection regulations and, further, the likely requirement to increase shielding thickness still more at a later date can be done usually by heaping more earth or applying local shielding at minimal expense and inconvenience. This moderately happy state of affairs does not unfortunately hold true with roof shielding. The cost of roof shielding is largely predicated on the roof span and the necessary structural engineering requirements for its support. These measures can be extremely expensive and where one is dealing with the rather extensive unsupported spans typical of experimental halls devoted to experiments with high energy electron beams; it is necessary to specify the roof thickness as carefully as possible with the constant concern that adding more earth later is not likely to be possible without rebuilding the hall. Because of the nature of roof skyshine, and for most high energy accelerator facilities neutron skyshine, the effect of the radiation is likely to extend to the facility fence-line where one is concerned about the exposure of the general population. Very properly the dose limit for the general population is set at a rather low value (1 mSv y -1 ) and in order for the Jefferson Lab (JLab) to ensure strict compliance with this limit they have a design goal for the fence line of 0.1 mSv y -1 . However, because natural neutron backgrounds are low (30--40 microSv y -1 ) and the methods of detection and measurement permit rejection of background interference from photons, they can measure the JLab produced neutron radiation with good sensitivity and precision

  10. Sodium fast reactor power monitoring using gamma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.; Normand, S.; Barbot, L.; Domenech, T.; Kondrasovs, V.; Corre, G.; Frelin, A.M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, CEA - Saclay DRT/LIST/DETECS/SSTM, Batiment 516 - P.C. no 72, Gif sur Yvette, F-91191 (France); Montagu, T.; Dautremer, T.; Barat, E. [CEA, LIST, Laboratoire Processus Stochastiques et Spectres (France); Ban, G. [ENSICAEN (France)

    2009-06-15

    This work deals with the use of high flux gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. The simulation study part of this work has shown that power monitoring in a short time response and with a good accuracy is possible. An experimental test is under preparation at the French SFR Phenix experimental reactor to validate simulation studies. First, physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as the sodium velocity, the atomic densities, Phenix neutron spectrum and incident neutron cross-sections of reactions producing gamma emitters. A thermal hydraulic transfer function was used for modeling primary sodium flow in our calculations. For the power monitoring problematic, use of a short decay period gamma emitter will allow to have a very fast response system without cumulative effect. We have determined that the best tagging agent is 20F which emits 1634 keV energy photons with a decay period of 11 s. The gamma spectrum was determined by flux point and a pulse high tally MCNP5.1.40 simulation and shown the possibility to measure the signal of this radionuclide. The experiment will be set during the reactor 'end life testing'. The Delayed Neutron Detection (DND) room has been chosen as the best available location on Phenix reactor to measure this kind of radionuclide due to a short transit time from reactor core to measurement sample. This location is optimum for global power measurement because homogenized sampling in the reactor hot pool. The main spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The HPGe diode signal will be processed by the Adonis digital signal processing due to high flux and fast activity measurement. Post-processing softwares will be used to limit statistical problems of the

  11. Explicit calculation of multi-fold contour integrals of certain ratios of Euler gamma functions. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Ivan [Valparaiso Univ. (Chile). Inst. de Fisica y Astronomia; Kniehl, Bernd A. [Hamburg Univ. (Germany). II. Inst. fuer Theoretische Physik; Kondrashuk, Igor [Univ. del Bio-Bio, Chillan (Chile). Dept. de Ciencias Basicas; Notte-Cuello, Eduardo A. [La Serena Univ. (Chile). Dept. de Matematicas; Parra-Ferrada, Ivan [Talca Univ. (Chile). Inst. de Matematica y Fisica; Rojas-Medar, Marko A. [Univ. de Tarapaca, Arica (Chile). Inst. de Alta Investigacion

    2016-12-15

    In this paper we proceed to study properties of Mellin-Barnes (MB) transforms of Usyukina-Davydychev (UD) functions. In our previous papers [Nuclear Physics B 870 (2013) 243], [Nuclear Physics B 876 (2013) 322] we showed that multi-fold Mellin-Barnes (MB) transforms of Usyukina-Davydychev (UD) functions may be reduced to two-fold MB transforms and that the higher-order UD functions were obtained in terms of a differential operator by applying it to a slightly modified first UD function. The result is valid in d=4 dimensions and its analog in d=4-2ε dimensions exits too [Theoretical and Mathematical Physics 177 (2013) 1515]. In [Nuclear Physics B 870 (2013) 243] the chain of recurrent relations for analytically regularized UD functions was obtained implicitly by comparing the left hand side and the right hand side of the diagrammatic relations between the diagrams with different loop orders. In turn, these diagrammatic relations were obtained due to the method of loop reductions for the triangle ladder diagrams proposed in 1983 by Belokurov and Usyukina. Here we reproduce these recurrent relations by calculating explicitly via Barnes lemmas the contour integrals produced by the left hand sides of the diagrammatic relations. In such a way we explicitly calculate a family of multi-fold contour integrals of certain ratios of Euler gamma functions. We make a conjecture that similar results for the contour integrals are valid for a wider family of smooth functions which includes the MB transforms of UD functions.

  12. LHCb Observation of photon polarization in the $b\\rightarrow s\\gamma$ transition

    CERN Multimedia

    Veneziano, Giovanni

    2014-01-01

    The Standard Model (SM) predicts that the photon emitted in $b\\rightarrow s\\gamma$ transitions is predominantly left-handed. While the measured inclusive $b\\rightarrow s\\gamma$ rate agrees with the SM calculations, no direct evidence exists for a nonzero photon polarization $\\lambda_\\gamma$ in this type of decays. Several extensions of the SM, compatible with all current measurements, predict that the photon acquires a significant right-handed component.

  13. Analysis of gamma irradiator dose rate using spent fuel elements with parallel configuration

    International Nuclear Information System (INIS)

    Setiyanto; Pudjijanto MS; Ardani

    2006-01-01

    To enhance the utilization of the RSG-GAS reactor spent fuel, the gamma irradiator using spent fuel elements as a gamma source is a suitable choice. This irradiator can be used for food sterilization and preservation. The first step before realization, it is necessary to determine the gamma dose rate theoretically. The assessment was realized for parallel configuration fuel elements with the irradiation space can be placed between fuel element series. This analysis of parallel model was choice to compare with the circle model and as long as possible to get more space for irradiation and to do manipulation of irradiation target. Dose rate calculation were done with MCNP, while the estimation of gamma activities of fuel element was realized by OREGEN code with 1 year of average delay time. The calculation result show that the gamma dose rate of parallel model decreased up to 50% relatively compared with the circle model, but the value still enough for sterilization and preservation. Especially for food preservation, this parallel model give more flexible, while the gamma dose rate can be adjusted to the irradiation needed. The conclusion of this assessment showed that the utilization of reactor spent fuels for gamma irradiator with parallel model give more advantage the circle model. (author)

  14. Recent progress in low-level gamma imaging

    International Nuclear Information System (INIS)

    Mahe, C.; Girones, Ph.; Lamadie, F.; Le Goaller, C.

    2007-01-01

    The CEA's Aladin gamma imaging system has been operated successfully for several years in nuclear plants and during decommissioning projects with additional tools such as gamma spectrometry detectors and dose rate probes. The radiological information supplied by these devices is becoming increasingly useful for establishing robust and optimized decommissioning scenarios. Recent technical improvements allow this gamma imaging system to be operated in low-level applications and with shorter acquisition times suitable for decommissioning projects. The compact portable system can be used in places inaccessible to operators. It is quick and easy to implement, notably for onsite component characterization. Feasibility trials and in situ measurements were recently carried out under low-level conditions, mainly on waste packages and glove boxes for decommissioning projects. This paper describes recent low-level in situ applications. These characterization campaigns mainly concerned gamma emitters with γ energy < 700 keV. In many cases, the localization of hot spots by gamma camera was confirmed by additional measurements such as dose rate mapping and gamma spectrometry measurements. These complementary techniques associated with advanced calculation codes (MCNP, Mercure 6.2, Visiplan and Siren) offer a mobile and compact tool for specific assessment of waste packages and glove boxes. (authors)

  15. Application of Inverse Gamma Transport to Material Thickness Identification with SGRD Code

    Directory of Open Access Journals (Sweden)

    Humbert Philippe

    2017-01-01

    Full Text Available SGRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used to infer the dimensions of a one dimensional model of a shielded gamma ray source. The method is based on the simulation of the uncollided leakage current of discrete gamma lines that are produced by nuclear decay. Experimentally, the unscattered gamma lines leakage current is obtained by processing high precision gamma spectroscopy measurements. The material thicknesses are computed with SGRD using a fast ray-tracing algorithm embedded in a non-linear multidimensional iterative optimization procedure that minimizes the error metric between calculated and measured signatures. For verification, numerical results on a test problem are presented.

  16. Efficiency of a bismuth-germanate scintillator: comparison of Monte Carlo calculations with measurements

    International Nuclear Information System (INIS)

    Hsu, H.H.; Dowdy, E.J.; Estes, G.P.; Lucas, M.C.; Mack, J.M.; Moss, C.E.; Hamm, M.E.

    1983-01-01

    Monte Carlo calculations of a bismuth-germanate scintillator's efficiency agree closely with experimental measurements. For this comparison, we studied the absolute gamma-ray photopeak efficiency of a scintillator (7.62 cm long by 7.62 cm in diameter) at several gamma-ray energies from 166 to 2615 keV at distances from 0.5 to 152.4 cm. Computer calculations were done in a two-dimensional cylindrical geometry with the Monte Carlo coupled photon-electron code CYLTRAN. For the experiment we measured 11 sources with simple spectra and precisely known strengths. The average deviation between the calculations and the measurements is 3%. Our calculated results also closely agree with recently published calculated results

  17. gamma. -ray. Present status and problems

    Energy Technology Data Exchange (ETDEWEB)

    Okudaira, K [Rikkyo Univ., Tokyo (Japan). Faculty of Science

    1975-01-01

    As ..gamma..-ray advances straightly through space, the study on cosmic ..gamma..-ray will give the information concerning the origin directly. However, the intensity is weak, and the avoidance of background is a serious problem. The wide-spread components were studied by OSO-3. The intensity of the galactic disc component around 100 MeV was reported as (3.4+-1.0)x10/sup -5/ photons (cm/sup 2/, radian, sec)/sup -1/ by OSO-3 and 0.2x10/sup -4/ photons (cm/sup 2/, radian sec)/sup -1/ by SAS-2, and corresponds to the calculated ..gamma.. yield from ..pi../sup 0/. The strong disc component, so-called galactic center region, has been observed, and is due to the mixture of ..gamma..-ray from ..pi../sup 0/ and inverse Compton ..gamma..-ray. A peak at 476+-24 KeV was found as well as the continuous component. Special care must be taken for the observation of isotropic component, since it is hardly distinguished from the background. It is considered that the isotropic component is due to the inverse Compton scattering of 3/sup 0/K radiation in super-galactic space and the contribution from outer galaxy. The nearest point source of ..gamma..-ray is the sun. Among the other point sources, the crab nebula is the most reliable one. The energy flux of pulse component showed the spectrum of E/sup -1/. ..gamma..-ray bursts were observed by man-made satellites Vela-5 and 6. Theoretical explanation is still incomplete regarding the bursts. (Kato, T.).

  18. Interpretation of the galactic radio-continuum and gamma-ray emission

    International Nuclear Information System (INIS)

    Beuermann, K.P.

    1974-01-01

    An analysis is performed of the nonthermal radio-continuum and gamma-ray emission of the galactic disc, using a spiral-arm model of the Galaxy. The results for the 408 MHz brightness temperature and the >100 MeV gamma-ray line intensity as a function of galactic longitude at bsup(II)=0 deg are presented. The observational implications, as well as the uncertainties in the calculations, are briefly discussed. An estimate of the possible range of the inverse Compton contribution to the observed gamma-ray flux is made

  19. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

  20. ECP evaluation by water radiolysis and ECP model calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

    2010-07-01

    In-pile ECP measurements data was evaluated by water radiolysis calculations. The data was obtained by using an in-pile loop in an experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Three types of ECP sensors, a Pt electrode, an Ag/AgCl sensor and a zirconia membrane sensor containing Ag/Ag{sub 2}O were used at several levels of the irradiation rig at various neutron flux and gamma rates. For water radiolysis calculation, the in-pile loop was modeled to several nodes following their design specifications, operating conditions such as flow rates, dose rate distributions of neutron and gamma-ray and so on. Concentration of chemical species along the water flow was calculated by a radiolysis code, WRAC-J. The radiolysis calculation results were transferred to an ECP model. In the model, anodic and cathodic current densities were calculated with combination of an electrochemistry model and an oxide film growth model. The measured ECP data were compared with the radiolysis/ECP calculation results, and applicability the of radiolysis model was confirmed. In addition, anomalous phenomenon appears in the in-pile loop was also investigated by radiolysis calculations. (author)

  1. Relativistic effects in gamma-ray bursts

    International Nuclear Information System (INIS)

    Eriksen, Erik; Groen, Oeyvind

    1999-01-01

    According to recent models of the sources of gamma-ray bursts the extremely energetic emission is caused by shells expanding with ultrarelativistic velocity. With the recent identification of optical sources at the positions of some gamma-ray bursts these ''fireball'' models have acquired an actuality that invites to use them as a motivating application when teaching special relativity. We demonstrate several relativistic effects associated with these models which are very pronounced due to the great velocity of the shell. For example a burst lasting for a month in the rest frame of an element of the shell lasts for a few seconds only, in the rest frame of our detector. It is shown how the observed properties of a burst are modified by aberration and the Doppler effect. The apparent luminosity as a function of time is calculated. Modifications due to the motion of the star away from the observer are calculated. (Author)

  2. Characteristics of environmental gamma-rays and dose assessment

    International Nuclear Information System (INIS)

    Saito, Kimiaki; Moriuchi, Shigeru

    1986-01-01

    Environmental radioactivity has attracted much attention in terms of exposure to the population, although its exposure doses are minimal. This paper presents problems encountered in the assessment of exposure doses using model and monitoring systems, focusing on the characteristics, such as energy distribution, direction distribution, and site, of environmental gamma-rays. The assessment of outdoor and indoor exposure doses of natural gamma-rays is discussed in relation to the shielding effect of the human body. In the assessment of artificial gamma-rays, calculation of exposure doses using build-up factor, the shielding effect of the human body, and energy dependency of the measuring instrument are covered. A continuing elucidation about uncertainties in dose assessment is emphasized. (Namekawa, K.)

  3. Hot particle dose calculations using the computer code VARSKIN Mod 2

    International Nuclear Information System (INIS)

    Durham, J.S.

    1991-01-01

    The only calculational model recognised by the Nuclear Regulatory Commission (NRC) for hot particle dosimetry is VARSKIN Mod 1. Because the code was designed to calculate skin dose from distributed skin contamination and not hot particles, it is assumed that the particle has no thickness and, therefore, that no self-absorption occurs within the source material. For low energy beta particles such as those emitted from 60 Co, a significant amount of self-shielding occurs in hot particles and VARSKIN Mod 1 overestimates the skin dose. In addition, the presence of protective clothing, which will reduce the calculated skin dose for both high and low energy beta emitters, is not modelled in VARSKIN Mod 1. Finally, there is no provision in VARSKIN Mod 1 to calculate the gamma contribution to skin dose from radionuclides that emit both beta and gamma radiation. The computer code VARSKIN Mod 1 has been modified to model three-dimensional sources, insertion of layers of protective clothing between the source and skin, and gamma dose from appropriate radionuclides. The new code, VARSKIN Mod 2, is described and the sensitivity of the calculated dose to source geometry, diameter, thickness, density, and protective clothing thickness are discussed. Finally, doses calculated using VARSKIN Mod 2 are compared to doses measured from hot particles found in nuclear power plants. (author)

  4. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  5. Shielding calculations for the SNO detector

    International Nuclear Information System (INIS)

    Earle, E.D.; Wong, P.Y.

    1987-05-01

    The gamma-ray background into the central D 2 O vessel of the SNO detector due to Th and U in the rock, concrete, and photomultipliers is calculated. A cylindrical geometry and concrete thicknesses of 0.5 and 1 m are assumed. The effect of adding boron to the concrete is also considered. It is concluded that backgrounds from (α,n) reactions can be reduced to the required level. These calculations will assist in finalizing the detector design but additional calculations will be required as new design details become known

  6. Interferences in Prompt {gamma} Analysis of corrosive contaminants in concrete

    Energy Technology Data Exchange (ETDEWEB)

    Naqvi, A.A. [Department of Physics, King Fahd University of Petroleum and Minerals, KFUPM Box 1815, Dhahran-31261 (Saudi Arabia)]. E-mail: aanaqvi@kfupm.edu.sa; Nagadi, M.M. [Department of Physics, King Fahd University of Petroleum and Minerals, KFUPM Box 1815, Dhahran-31261 (Saudi Arabia); Al-Amoudi, O.S.B. [Department of Civil Engineering, King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia)

    2006-12-21

    An accelerator-based Prompt Gamma Neutron Activation Analysis (PGNAA) setup has been developed to measure the concentration of corrosive chloride and sulfate contaminants in concrete. The Minimum Detectable Concentration (MDC) limit of chlorine and sulfur in the concrete depends upon the {gamma}-ray used for elemental analysis. For more interfering {gamma}-rays, the MDC limit is higher than that for less interfering {gamma}-rays. The MDC limit of sulfur in concrete measured for the KFUPM PGNAA setup was calculated to be 0.60{+-}0.19 wt%. The MDC limit is equal to the upper limit of sulfur concentration in concrete set by the British Standards. The MDC limit of chlorine in concrete for the KFUPM PGNAA setup, which was calculated for less interfering 1.165 MeV {gamma}-rays, was found to be 0.075{+-}0.025 wt%. The lower limits of the MDC of chlorine in concrete was 73% higher than the limit set by American Concrete Institute. The limit of the MDC can be improved to the desired standard by increasing the intensity of neutron source. For moreinterfering 5.715 and 6.110 MeV chlorine {gamma}-rays the MDC limit was found to be 2-3 times larger than that of 1.165 MeV {gamma}-rays. When normalized to the same intensity of the neutron source, the MDC limits of chlorine and sulfur in concrete from the KFUPM PGNAA setup are better than MDC limits of chlorine in concrete obtained with the {sup 241}Am-Be source-based PGNAA setup. This study has shown that an accelerator-based PGNAA setup can be used in chlorine and sulfur analysis of concrete samples.

  7. EELOSS: the program for calculation of electron energy loss data

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi

    1980-10-01

    A computer code EELOSS has been developed to obtain the electron energy loss data required for shielding and dosimetry of beta- and gamma-rays in nuclear plants. With this code, the following data are obtainable for any energy from 0.01 to 15 MeV in any medium (metal, insulator, gas, compound, or mixture) composed of any choice of 69 elements with atomic number 1 -- 94: a) Collision stopping power, b) Restricted collision stopping power, c) Radiative stopping power, and d) Bremsstrahlung production cross section. The availability of bremsstrahlung production cross section data obtained by the EELOSS code is demonstrated by the comparison of calculated gamma-ray spectrum with measured one in Pb layer, where electron-photon cascade is included implicitly. As a result, it is concluded that the uncertainty in the bremsstrahlung production cross sections is negligible in the practical shielding calculations of gamma rays of energy less than 15 MeV, since the bremsstrahlung production cross sections increase with the gamma-ray energy and the uncertainty for them decreases with increasing the gamma-ray energy. Furthermore, the accuracy of output data of the EELOSS code is evaluated in comparison with experimental data, and satisfactory agreements are observed concerning the stopping power. (J.P.N.)

  8. Attenuation of Reactor Gamma Radiation and Fast Neutrons Through Large Single-Crystal Materials

    International Nuclear Information System (INIS)

    Adib, M.

    2009-01-01

    A generalized formula is given which, for neutron energies in the range 10-4< E< 10 eV and gamma rays with average energy 2 MeV , permits calculation of the transmission properties of several single crystal materials important for neutron scattering instrumentation. A computer program Filter was developed which permits the calculation of attenuation of gamma radiation, nuclear capture, thermal diffuse and Bragg-scattering cross-sections as a function of materials constants, temperature and neutron energy. The applicability of the deduced formula along with the code checked from the obtained agreement between the calculated and experimental neutron transmission through various single-crystals A feasibility study for use of Si, Ge, Pb, Bi and sapphire is detailed in terms of optimum crystal thickness, mosaic spread and cutting plane for efficient transmission of thermal reactor neutrons and for rejection of the accompanying fast neutrons and gamma rays.

  9. Using gamma distribution to determine half-life of rotenone, applied in freshwater

    Energy Technology Data Exchange (ETDEWEB)

    Rohan, Maheswaran, E-mail: mrohan@aut.ac.nz [Department of Biostatistics and Epidemiology, Auckland University of Technology, Auckland (New Zealand); Fairweather, Alastair; Grainger, Natasha [Science and Capability, Department of Conservation, Hamilton (New Zealand)

    2015-09-15

    Following the use of rotenone to eradicate invasive pest fish, a dynamic first-order kinetic model is usually used to determine the half-life and rate at which rotenone dissipated from the treated waterbody. In this study, we investigate the use of a stochastic gamma model for determining the half-life and rate at which rotenone dissipates from waterbodies. The first-order kinetic and gamma models produced similar values for the half-life (4.45 days and 5.33 days respectively) and days to complete dissipation (51.2 days and 52.48 days respectively). However, the gamma model fitted the data better and was more flexible than the first-order kinetic model, allowing us to use covariates and to predict a possible range for the half-life of rotenone. These benefits are particularly important when examining the influence that different environmental factors have on rotenone dissipation and when trying to predict the rate at which rotenone will dissipate during future operations. We therefore recommend that in future the gamma distribution model is used when calculating the half-life of rotenone in preference to the dynamic first-order kinetics model. - Highlights: • We investigated the use of the gamma model to calculate the half-life of rotenone. • Physical and environmental variables can be incorporated into the model. • A method for calculating the range around a mean half-life is presented. • The model is more flexible than the traditionally used first-order kinetic model.

  10. Estimation of neutron energy distributions from prompt gamma emissions

    Science.gov (United States)

    Panikkath, Priyada; Udupi, Ashwini; Sarkar, P. K.

    2017-11-01

    A technique of estimating the incident neutron energy distribution from emitted prompt gamma intensities from a system exposed to neutrons is presented. The emitted prompt gamma intensities or the measured photo peaks in a gamma detector are related to the incident neutron energy distribution through a convolution of the response of the system generating the prompt gammas to mono-energetic neutrons. Presently, the system studied is a cylinder of high density polyethylene (HDPE) placed inside another cylinder of borated HDPE (BHDPE) having an outer Pb-cover and exposed to neutrons. The emitted five prompt gamma peaks from hydrogen, boron, carbon and lead can be utilized to unfold the incident neutron energy distribution as an under-determined deconvolution problem. Such an under-determined set of equations are solved using the genetic algorithm based Monte Carlo de-convolution code GAMCD. Feasibility of the proposed technique is demonstrated theoretically using the Monte Carlo calculated response matrix and intensities of emitted prompt gammas from the Pb-covered BHDPE-HDPE system in the case of several incident neutron spectra spanning different energy ranges.

  11. Gamma sensitivity of the Eberline PCM-1

    International Nuclear Information System (INIS)

    Blanton, J.D.

    1988-01-01

    This paper reports that normally, alarm setpoints for the Eberline PCM-1 series personnel contamination monitors are calculated based upon efficiencies measured using 100-cm 2 or larger beta or mixed beta-gamma sources. This simulates the type of contamination most frequently encountered on personnel and clothing--low-level, distributed beta-gamma emitters. In most circumstances, the PCM-1's sensitivity to the other type of contamination encountered in nuclear plant work--hot particles--would be expected to be the same as, or better than, its sensitivity to distributed contamination. However, particles that are deposited on skin or clothing can be partially shielded from the view of the PCM-1's detectors. In these situations, the PCm-1's sensitivity to gamma radiation may be more relevant than its sensitivity to betas

  12. Direct transitions from high-K isomers to low-K bands -- {gamma} softness or coriolis coupling

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Yoshifumi R.; Narimatsu, Kanako; Ohtsubo, Shin-Ichi [Kyushu Univ., Fukuoka (Japan)] [and others

    1996-12-31

    Recent measurements of direct transitions from high-K isomers to low-K bands reveal severe break-down of the K-selection rule and pose the problem of how to understand the mechanism of such K-violation. The authors recent systematic calculations by using a simple {gamma}-tunneling model reproduced many of the observed hindrances, indicating the importance of the {gamma} softness. However, there are some data which cannot be explained in terms of the {gamma}-degree of freedom. In this talk, the authors also discuss the results of conventional Coriolis coupling calculations, which is considered to be another important mechanism.

  13. Dose Calculation Accuracy of the Monte Carlo Algorithm for CyberKnife Compared with Other Commercially Available Dose Calculation Algorithms

    International Nuclear Information System (INIS)

    Sharma, Subhash; Ott, Joseph; Williams, Jamone; Dickow, Danny

    2011-01-01

    Monte Carlo dose calculation algorithms have the potential for greater accuracy than traditional model-based algorithms. This enhanced accuracy is particularly evident in regions of lateral scatter disequilibrium, which can develop during treatments incorporating small field sizes and low-density tissue. A heterogeneous slab phantom was used to evaluate the accuracy of several commercially available dose calculation algorithms, including Monte Carlo dose calculation for CyberKnife, Analytical Anisotropic Algorithm and Pencil Beam convolution for the Eclipse planning system, and convolution-superposition for the Xio planning system. The phantom accommodated slabs of varying density; comparisons between planned and measured dose distributions were accomplished with radiochromic film. The Monte Carlo algorithm provided the most accurate comparison between planned and measured dose distributions. In each phantom irradiation, the Monte Carlo predictions resulted in gamma analysis comparisons >97%, using acceptance criteria of 3% dose and 3-mm distance to agreement. In general, the gamma analysis comparisons for the other algorithms were <95%. The Monte Carlo dose calculation algorithm for CyberKnife provides more accurate dose distribution calculations in regions of lateral electron disequilibrium than commercially available model-based algorithms. This is primarily because of the ability of Monte Carlo algorithms to implicitly account for tissue heterogeneities, density scaling functions; and/or effective depth correction factors are not required.

  14. Detecting gamma-ray anisotropies from decaying dark matter. Prospects for Fermi LAT

    International Nuclear Information System (INIS)

    Ibarra, Alejandro; Tran, David

    2009-09-01

    Decaying dark matter particles could be indirectly detected as an excess over a simple power law in the energy spectrum of the diffuse extragalactic gamma-ray background. Furthermore, since the Earth is not located at the center of the Galactic dark matter halo, the exotic contribution from dark matter decay to the diffuse gamma-ray flux is expected to be anisotropic, offering a complementary method for the indirect search for decaying dark matter particles. In this paper we discuss in detail the expected dipole-like anisotropies in the dark matter signal, taking also into account the radiation from inverse Compton scattering of electrons and positrons from dark matter decay. A different source for anisotropies in the gamma-ray flux are the dark matter density fluctuations on cosmic scales. We calculate the corresponding angular power spectrum of the gamma-ray flux and comment on observational prospects. Finally, we calculate the expected anisotropies for the decaying dark matter scenarios that can reproduce the electron/positron excesses reported by PAMELA and the Fermi LAT, and we estimate the prospects for detecting the predicted gamma-ray anisotropy in the near future. (orig.)

  15. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

    1997-10-01

    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  16. Density functional study of gamma-aminopropyltriethoxysilane

    International Nuclear Information System (INIS)

    Bistricic, L; Volovsek, V; Daani, V; Leskovac, M

    2006-01-01

    Density functional theory calculations using Becke's three-parameter exchange functional in combination with the Lee-Young-Parr correlation functional (B3-LYP) and standard 6-311 + G(d,p) basis set were carried out to study the conformational stability and vibrational spectra of gamma-aminopropyltriethoxysilane. Calculations reveal the existence of two stable conformers trans and gauche. The calculated energy for the gauche conformation was found to be 608 cm -1 above the minimum energy of the trans conformation. Temperature dependence of Raman spectra of liquid APTES and DFT calculation enabled us to identify the vibrational bands characteristic for both conformers. It has been shown that there is an increase in the population of gauche conformer with increasing temperature

  17. Continued Development of a Soft Gamma-Ray Concentrator

    Science.gov (United States)

    Bloser, Peter

    We propose to continue our development of a concept for a soft gamma-ray (E > 100 keV) concentrator using thin-film multilayer structures. Alternating layers of low- and high-density materials will channel soft gamma-ray photons via total external reflection. A suitable arrangement of bent structures will then concentrate the incident radiation to a point. Gamma-ray optics made in this way offer the potential for soft gamma-ray telescopes with focal lengths of less than 10 m, removing the need for formation flying spacecraft and opening the field up to balloon-borne instruments. Under previous APRA funding we have been investigating methods for efficiently producing such multilayer structures and modeling their performance. We now propose to pursue magnetron sputtering (MS) techniques to quickly produce structures with the required smoothness and thickness, to measure their channeling efficiency and compare with calculations, and to design a "lens" with optimized bandpass and throughput and predict its scientific performance. If successful, this work will confirm that this innovative optics concept is suitable for a balloon-born soft gamma-ray telescope with unprecedented sensitivity.

  18. A novel dual mode neutron-gamma imager

    International Nuclear Information System (INIS)

    Cooper, Robert Lee; Gerling, Mark; Brennan, James S.; Mascarenhas, Nicholas; Mrowka, Stanley; Marleau, Peter

    2010-01-01

    The Neutron Scatter Camera (NSC) can image fission sources and determine their energy spectra at distances of tens of meters and through significant thicknesses of intervening materials in relatively short times (1). We recently completed a 32 element scatter camera and will present recent advances made with this instrument. A novel capability for the scatter camera is dual mode imaging. In normal neutron imaging mode we identify and image neutron events using pulse shape discrimination (PSD) and time of flight in liquid scintillator. Similarly gamma rays are identified from Compton scatter in the front and rear planes for our segmented detector. Rather than reject these events, we show it is possible to construct a gamma-ray image by running the analysis in a 'Compton mode'. Instead of calculating the scattering angle by the kinematics of elastic scatters as is appropriate for neutron events, it can be found by the kinematics of Compton scatters. Our scatter camera has not been optimized as a Compton gamma-ray imager but is found to work reasonably. We studied imaging performance using a Cs137 source. We find that we are able to image the gamma source with reasonable fidelity. We are able to determine gamma energy after some reasonable assumptions. We will detail the various algorithms we have developed for gamma image reconstruction. We will outline areas for improvement, include additional results and compare neutron and gamma mode imaging.

  19. Using MCBEND for neutron or gamma-ray deterministic calculations

    Science.gov (United States)

    Geoff, Dobson; Adam, Bird; Brendan, Tollit; Paul, Smith

    2017-09-01

    MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler's ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.

  20. Gamma rays from Cygnus X-1: Modeling and nonthermal pair production

    International Nuclear Information System (INIS)

    Dermer, C.D.; Liang, E.P.

    1988-02-01

    The gamma-ray bump observed between 0.5 and 2 MeV in the spectrum of Cygnus X-1 can be interpreted as the thermal emissions from a hot (kT/approximately/400 keV) pair-dominated cloud. We argue that the X-rays and gamma rays are produced in separate emission regions, and calculate the photon-photon pair production rate from X-ray and gamma-ray interactions in the vicinity of Cyg X-1 by employing a simplified geometry for the two emitting regions

  1. Gamma radiation in apartments

    International Nuclear Information System (INIS)

    Grindborg, J.-E.

    1983-05-01

    This investigation forms the basis for the description of methods for the detection of gamma radiation. The aim is to control that the dose limit will not exceed 50 μR/h in a room where people reside. The distribution of dose rates in different rooms has been calculated and the results have been compared with experimental data. Various instruments have been calibrated and their specifications are discussed. (G.B.)

  2. Skyshine analysis using various nuclear data files

    International Nuclear Information System (INIS)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Nomura, Y.; Tsubosaka, A.

    2000-01-01

    The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

  3. Evaluation of the neutron and gamma-ray production cross-sections for 55Mn

    International Nuclear Information System (INIS)

    Takahashi, H.

    1974-11-01

    The evaluation of neutron and gamma production cross sections for manganese-55 from 1.0 (10) -5 eV to 20.0 MeV for ENDF/ B-IV is summarized. Included are resonance parameters, neutron cross sections, angular and energy distribution of secondary neutrons, gamma multiplicities and transition probability array, gamma angular and energy distributions, nuclear model calculations, uncertainty estimates of cross sections, and evaluated cross sections. (U.S.)

  4. Measurements of neutron and gamma ray streaming through a duct, (2), (3)

    International Nuclear Information System (INIS)

    Hashikura, Hiroyuki; Fukumoto, Hideshi; Akiyama, Masatsugu; Oka, Yoshiaki; An, Shigehiro

    1982-03-01

    Measurements of neutron and gamma ray streaming through a duct measurements of and a cavity in concrete shields were measured in the fast neutron source reactor YAYOI of the University of Tokyo. The neutron spectra measured by a NE213 scintillator and proton recoil counters were compared with the calculations using Monte Carlo code, MORSE-CG. The agreements between the experiments and the calculations were generally satisfactory. The attenuations of neutron and gamma ray in the cavity and the duct were studied in the three experimental configurations. (author)

  5. Determination of correction and conversion factor of exposure rate generated Gamma spectrometer GR-320 to Victoreen data

    International Nuclear Information System (INIS)

    Supardjo-AS; Mappa, Djody-Rachim; Nasrun-Syamsul; Syamsul-Hadi

    2000-01-01

    Exposure rate data of Muria Peninsula were generated from Victoreen-491 measurement and calculation of radioelement content in soil which were measured by Exploranium GR-320, using IAEA formula. However those data are not be comparable so the exposure rate calculated from Gamma Spectrometer data necessarily to be corrected. The correction factor was determinate by measuring the exposure rate of at the NMDC's back yard selected location using Victoreen-491 and Gamma Spectrometer Exploranium GR-320 . Correction factor was created by comparing mean exposure rate data that calculated from 30 data measured by Gamma Spectrometer instrument and to those Victoreen's exposure rate. Conversion factor was gained from comparing of total count data of Gamma Spectrometer Exploranium GR-320 to Victoreen's exposure rate data. The correction factor of Exploranium GR-320's exposure rate is 0.34 μR/hours, and the conversion factor of total count is 0.0092 μR/hours per c/m. Deviation Victoreen 491 = 4.7 % and Gamma Spectrometer Exploranium GR-320 8.6 %

  6. Gamma-spectrometry of extended sources for analysing environmental samples

    International Nuclear Information System (INIS)

    Jarosievitz, B.

    1996-01-01

    Measurements of the environmental activity concentration by gamma spectrometers require the determination of the full-energy-peak efficiency as a function of photon energy over the detector range. This can be done by experiments or by calculation. For simple cases, experiments are straightforward, but if the decay scheme is complex, cascade effects modify detection efficiency. Also, actual detection efficiency depends on the detection geometry. All these effects are treated as corrections or modifications of the simple value cases which are especially relevant when applied to large volume of environmental samples. In this thesis calculations are made, using the GEANT MC program, for realistic experimental situations that have been performed, and these calculations are validated. The calculational and experimental results have been compared, and if it proves to be satisfactory, the results can be relied on even for cases when no direct experimental observation is possible. The general problems of gamma spectroscopy and correction problems are discussed. The two main tools, the experimental setup and the simulation program are described. A careful checking of the simulation results and the consequences are presented. (R.P.)

  7. Asymmetry of the cross section for the reaction. gamma. d. -->. pi. /sup 0/d with linearly polarized. gamma. rays at 500--700 MeV and at a c. m. angle theta(0 = 130/sup 0/

    Energy Technology Data Exchange (ETDEWEB)

    Adamyan, F.V.; Akopyan, G.G.; Vartapetyan, G.A.; Galumyan, P.I.; Grabskii, V.O.; Karapetyan, V.V.; Karapetyan, G.V.; Oktanyan, V.K.

    1984-06-25

    The asymmetry of the cross section (..sigma..) of the reaction ..gamma..d ..-->.. ..pi../sup 0/d induced by linearly polarized ..gamma.. rays has been measured at energies E..gamma.. = 500 MeV, E..gamma.. = 600, and E/sub ..gamma../ = 700 MeV at the c.m. angle theta(0 = 130/sup 0/. The results disagree with calculations in the impulse approximation. The results can be explained in a qualitative way by appealing to an /sup 3/F/sub 3/ (2.26-GeV) dibaryon resonance.

  8. Radiation effect on silicon transistors in mixed neutrons-gamma environment

    Science.gov (United States)

    Assaf, J.; Shweikani, R.; Ghazi, N.

    2014-10-01

    The effects of gamma and neutron irradiations on two different types of transistors, Junction Field Effect Transistor (JFET) and Bipolar Junction Transistor (BJT), were investigated. Irradiation was performed using a Syrian research reactor (RR) (Miniature Neutron Source Reactor (MNSR)) and a gamma source (Co-60 cell). For RR irradiation, MCNP code was used to calculate the absorbed dose received by the transistors. The experimental results showed an overall decrease in the gain factors of the transistors after irradiation, and the JFETs were more resistant to the effects of radiation than BJTs. The effect of RR irradiation was also greater than that of gamma source for the same dose, which could be because neutrons could cause more damage than gamma irradiation.

  9. Gamma ray shielding: a web based interactive program

    International Nuclear Information System (INIS)

    Subbaiah, K.V.; Senthi Kumar, C.; Sarangapani, R.

    2005-01-01

    A web based interactive computing program is developed using java for quick assessment of Gamma Ray shielding problems. The program addresses usually encountered source geometries like POINT, LINE, CYLINDRICAL, ANNULAR, SPHERICAL, BOX, followed by 'SLAB' shield configurations. The calculation is based on point kernel technique. The source points are randomly sampled within the source volume. From each source point, optical path traversed in the source and shield media up to the detector location is estimated to calculate geometrical and material attenuations, and then corresponding buildup factor is obtained, which accounts for scattered contribution. Finally, the dose rate for entire source is obtained by summing over all sampled points. The application allows the user to select one of the seven regular geometrical bodies and provision exist to give source details such as emission energies, intensities, physical dimensions and material composition. Similar provision is provided to specify shield slab details. To aid the user, atomic numbers, densities, standard build factor materials and isotope list with respective emission energies and intensity for ready reference are given in dropdown combo boxes. Typical results obtained from this program are validated against existing point kernel gamma ray shielding codes. Additional facility is provided to compute fission product gamma ray source strengths based on the fuel type, burn up and cooling time. Plots of Fission product gamma ray source strengths, Gamma ray cross-sections and buildup factors can be optionally obtained, which enable the user to draw inference on the computed results. It is expected that this tool will be handy to all health physicists and radiological safety officers as it will be available on the internet. (author)

  10. High energy neutrinos from gamma-ray bursts with precursor supernovae.

    Science.gov (United States)

    Razzaque, Soebur; Mészáros, Peter; Waxman, Eli

    2003-06-20

    The high energy neutrino signature from proton-proton and photo-meson interactions in a supernova remnant shell ejected prior to a gamma-ray burst provides a test for the precursor supernova, or supranova, model of gamma-ray bursts. Protons in the supernova remnant shell and photons entrapped from a supernova explosion or a pulsar wind from a fast-rotating neutron star remnant provide ample targets for protons escaping the internal shocks of the gamma-ray burst to interact and produce high energy neutrinos. We calculate the expected neutrino fluxes, which can be detected by current and future experiments.

  11. Characteristics of liver tissue for attenuate the gamma radiation

    International Nuclear Information System (INIS)

    Arcos P, A.; Rodriguez N, S.; Pinedo S, A.; Amador V, P.; Chacon R, A.; Vega C, H.R.

    2005-01-01

    It was determined the lineal attenuation coefficient of hepatic tissue before gamma radiation of a source of 137 Cs. When exposing organic material before X or gamma radiation fields, part of the energy of the photons is absorbed by the material, while another part crosses it without producing any effect. The quantity of energy that is absorbed is a measure of the dose that receives the material. The three main mechanisms by means of which the gamma rays interacting with the matter are: The Photoelectric Effect, the Compton dispersion and the Even production; the sum of these three processes is translated in the attenuation coefficient of the radiation. In this work we have used hepatic tissue of bovine, as substitute of the human hepatic tissue, and we have measured the lineal attenuation coefficient for photons of 0.662 MeV. Through a series of calculations we have determined the lineal attenuation coefficient for photons from 10 -3 to 10 -5 MeV and the measured coefficient was compared with the one calculated. (Author)

  12. Study The Validity of The Direct Mathematical Method For Calculation The Total Efficiency Using Point And Disk Sources

    International Nuclear Information System (INIS)

    Hagag, O.M.; Nafee, S.S.; Naeem, M.A.; El Khatib, A.M.

    2011-01-01

    The direct mathematical method has been developed for calculating the total efficiency of many cylindrical gamma detectors, especially HPGe and NaI detector. Different source geometries are considered (point and disk). Further into account is taken of gamma attenuation from detector window or any interfacing absorbing layer. Results are compared with published experimental data to study the validity of the direct mathematical method to calculate total efficiency for any gamma detector size.

  13. Thermodynamic calculation of the Fe-Zn-Si system

    Energy Technology Data Exchange (ETDEWEB)

    Su Xuping [Institute of Materials Research, School of Mechanical Engineering, Xiangtan University, Xiangtan 411105, Hunan (China)]. E-mail: sxping@xtu.edu.cn; Yin Fucheng [Institute of Materials Research, School of Mechanical Engineering, Xiangtan University, Xiangtan 411105, Hunan (China); Li Zhi [Institute of Materials Research, School of Mechanical Engineering, Xiangtan University, Xiangtan 411105, Hunan (China); Tang, N.-Y. [Teck Cominco Metals Ltd., Product Technology Centre, Mississauga, Ont., L5K 1B4 (Canada); Zhao Manxiu [Institute of Materials Research, School of Mechanical Engineering, Xiangtan University, Xiangtan 411105, Hunan (China)

    2005-06-21

    Silicon in steel significantly affects alloy growth kinetics in the coating in general galvanizing, thereby changing the coating microstructure from the usual stratified Fe-Zn alloy layers to a mass of {zeta} crystallites surrounding by liquid zinc. The Zn-Fe-Si phase diagram and the relevant thermodynamic information have great importance for the galvanizing industry in developing remedies for this problem. In this work, the available information on the Fe-Zn-Si system, including all three binary systems was reviewed and re-evaluated, and ternary parameters were extracted from the available experimental data. By assuming all the binary intermetallic phases with the exception of the {delta}, {gamma}{sub 1}, and {gamma} phases, have no ternary solubility, a thermodynamic calculation of the Fe-Zn-Si system was carried out, and relevant isothermal and isopleths sections were calculated. Its applicability in galvanizing industry was discussed. There is a good agreement between the calculated and the experimentally determined phase boundaries.

  14. $\\gamma$-Ray Pulsars: Emission Zones and Viewing Geometries

    OpenAIRE

    Romani, Roger W.; Yadigaroglu, I. -A.

    1994-01-01

    There are now a half dozen young pulsars detected in high energy photons by the Compton GRO, showing a variety of emission efficiencies and pulse profiles. We present here a calculation of the pattern of high energy emission on the sky in a model which posits $\\gamma$-ray production by charge depleted gaps in the outer magnetosphere. This model accounts for the radio to $\\gamma$-ray pulse offsets of the known pulsars, as well as the shape of the high energy pulse profiles. We also show that $...

  15. Accurate calculations of the WIMP halo around the Sun and prospects for its gamma-ray detection

    International Nuclear Information System (INIS)

    Sivertsson, Sofia; Edsjoe, Joakim

    2010-01-01

    Galactic weakly interacting massive particles (WIMPs) may scatter off solar nuclei to orbits gravitationally bound to the Sun. Once bound, the WIMPs continue to lose energy by repeated scatters in the Sun, eventually leading to complete entrapment in the solar interior. While the density of the bound population is highest at the center of the Sun, the only observable signature of WIMP annihilations inside the Sun is neutrinos. It has been previously suggested that although the density of WIMPs just outside the Sun is lower than deep inside, gamma rays from WIMP annihilation just outside the surface of the Sun, in the so-called WIMP halo around the Sun, may be more easily detected. We here revisit this problem using detailed Monte Carlo simulations and detailed composition and structure information about the Sun to estimate the size of the gamma-ray flux. Compared to earlier simpler estimates, we find that the gamma-ray flux from WIMP annihilations in the solar WIMP halo would be negligible; no current or planned detectors would be able to detect this flux.

  16. The application of an eddy diffusivity model to the dispersion of radionuclides in the atmosphere and the calculation of cloud gamma exposure

    International Nuclear Information System (INIS)

    Maul, P.R.

    1981-05-01

    A model which has been applied successfully to the study of the mesoscale transport of sulphur compounds can be adapted for radionuclides released from nuclear power stations. Although more complicated than the conventional Gaussian plume models it has several important advantages including the better representation of dry deposition and the variation of dispersion parameters with height above the surface. Building entrainment can be included in a straightforward manner and an approximate method can be used to incorporate isotope-dependent deposition velocities. A new method of calculating cloud gamma exposure is described which is particularly suited to eddy diffusivity models. This model will be used as an alternative to Gaussian plume methods in the BNL safety code NECTAR. (author)

  17. Using MCBEND for neutron or gamma-ray deterministic calculations

    Directory of Open Access Journals (Sweden)

    Geoff Dobson

    2017-01-01

    Full Text Available MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler’s ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.

  18. Skyshine analysis using various nuclear data files

    Energy Technology Data Exchange (ETDEWEB)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

  19. Calibration of Ge gamma-ray spectrometers for complex sample geometries and matrices

    Energy Technology Data Exchange (ETDEWEB)

    Semkow, T.M., E-mail: thomas.semkow@health.ny.gov [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Department of Environmental Health Sciences, School of Public Health, University at Albany, State University of New York, Rensselaer, NY 12144 (United States); Bradt, C.J.; Beach, S.E.; Haines, D.K.; Khan, A.J.; Bari, A.; Torres, M.A.; Marrantino, J.C.; Syed, U.-F. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Kitto, M.E. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Department of Environmental Health Sciences, School of Public Health, University at Albany, State University of New York, Rensselaer, NY 12144 (United States); Hoffman, T.J. [Wadsworth Center, New York State Department of Health, Empire State Plaza, Albany, NY 12201 (United States); Curtis, P. [Kiltel Systems, Inc., Clyde Hill, WA 98004 (United States)

    2015-11-01

    A comprehensive study of the efficiency calibration and calibration verification of Ge gamma-ray spectrometers was performed using semi-empirical, computational Monte-Carlo (MC), and transfer methods. The aim of this study was to evaluate the accuracy of the quantification of gamma-emitting radionuclides in complex matrices normally encountered in environmental and food samples. A wide range of gamma energies from 59.5 to 1836.0 keV and geometries from a 10-mL jar to 1.4-L Marinelli beaker were studied on four Ge spectrometers with the relative efficiencies between 102% and 140%. Density and coincidence summing corrections were applied. Innovative techniques were developed for the preparation of artificial complex matrices from materials such as acidified water, polystyrene, ethanol, sugar, and sand, resulting in the densities ranging from 0.3655 to 2.164 g cm{sup −3}. They were spiked with gamma activity traceable to international standards and used for calibration verifications. A quantitative method of tuning MC calculations to experiment was developed based on a multidimensional chi-square paraboloid. - Highlights: • Preparation and spiking of traceable complex matrices in extended geometries. • Calibration of Ge gamma spectrometers for complex matrices. • Verification of gamma calibrations. • Comparison of semi-empirical, computational Monte Carlo, and transfer methods of Ge calibration. • Tuning of Monte Carlo calculations using a multidimensional paraboloid.

  20. Calibration of Ge gamma-ray spectrometers for complex sample geometries and matrices

    International Nuclear Information System (INIS)

    Semkow, T.M.; Bradt, C.J.; Beach, S.E.; Haines, D.K.; Khan, A.J.; Bari, A.; Torres, M.A.; Marrantino, J.C.; Syed, U.-F.; Kitto, M.E.; Hoffman, T.J.; Curtis, P.

    2015-01-01

    A comprehensive study of the efficiency calibration and calibration verification of Ge gamma-ray spectrometers was performed using semi-empirical, computational Monte-Carlo (MC), and transfer methods. The aim of this study was to evaluate the accuracy of the quantification of gamma-emitting radionuclides in complex matrices normally encountered in environmental and food samples. A wide range of gamma energies from 59.5 to 1836.0 keV and geometries from a 10-mL jar to 1.4-L Marinelli beaker were studied on four Ge spectrometers with the relative efficiencies between 102% and 140%. Density and coincidence summing corrections were applied. Innovative techniques were developed for the preparation of artificial complex matrices from materials such as acidified water, polystyrene, ethanol, sugar, and sand, resulting in the densities ranging from 0.3655 to 2.164 g cm −3 . They were spiked with gamma activity traceable to international standards and used for calibration verifications. A quantitative method of tuning MC calculations to experiment was developed based on a multidimensional chi-square paraboloid. - Highlights: • Preparation and spiking of traceable complex matrices in extended geometries. • Calibration of Ge gamma spectrometers for complex matrices. • Verification of gamma calibrations. • Comparison of semi-empirical, computational Monte Carlo, and transfer methods of Ge calibration. • Tuning of Monte Carlo calculations using a multidimensional paraboloid

  1. Activity computer program for calculating ion irradiation activation

    Science.gov (United States)

    Palmer, Ben; Connolly, Brian; Read, Mark

    2017-07-01

    A computer program, Activity, was developed to predict the activity and gamma lines of materials irradiated with an ion beam. It uses the TENDL (Koning and Rochman, 2012) [1] proton reaction cross section database, the Stopping and Range of Ions in Matter (SRIM) (Biersack et al., 2010) code, a Nuclear Data Services (NDS) radioactive decay database (Sonzogni, 2006) [2] and an ENDF gamma decay database (Herman and Chadwick, 2006) [3]. An extended version of Bateman's equation is used to calculate the activity at time t, and this equation is solved analytically, with the option to also solve by numeric inverse Laplace Transform as a failsafe. The program outputs the expected activity and gamma lines of the activated material.

  2. Neutron and gamma-ray emission double differential cross sections for the nuclear reaction by 1.5 GeV {pi}{sup +} incidence

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori; Ishibashi, Kenji; Shigyo, Nobuhiro [Kyushu Univ., Fukuoka (Japan)] [and others

    1998-03-01

    Neutron and gamma-ray production double differential cross sections were measured for iron by the use of 1.5 GeV {pi}{sup +} mesons. The measured cross sections were compared with the calculated values by HETC-KFA2. For the neutrons, the calculated results deviate from the experimental data in the neutron energy region below 30 MeV. The calculated values of gamma-ray production agree with the experimental data at gamma-ray energies from 1 to 7 MeV within a factor of three. (author)

  3. Calculation of the external dose rate in the spent fuel pool for the case to use compact racks

    International Nuclear Information System (INIS)

    Passos, E.M. dos; Alves, A.S.M.

    1988-01-01

    The possible introduction of compact racks in the spent fuel pool of the Angra 1 Nuclear Power Plant largely inreases its storage capacity, but originates an increase of the gamma radiation sources. The precise evaluation of the effects of the adoption of this option on the external gamma dose rates and also on the thickness of the concrete shielding requires the utilization of sofisticated computer codes (QAD, ANISN), which allow the calculation of the gamma dose rates through thick shielding walls. This paper describes the utilized methodology for the calculation of the modified pool shieldings, showing the obtained results for the Angra 1 NPP case. The gamma dose rate was calculated with the point Kernel model, first analytically, and later through utilization of the tridimensional multigroup QAD computer code. (author) [pt

  4. Plutonium isotopic measurements by gamma-ray spectroscopy

    International Nuclear Information System (INIS)

    Haas, F.X.; Lemming, J.F.

    1976-01-01

    A nondestructive technique is described for calculating plutonium-238, plutonium-240, plutonium-241 and americium-241 relative to plutonium-239 from measured peak areas in the high resolution gamma-ray spectra of solid plutonium samples. Gamma-ray attenuation effects were minimized by selecting sets of neighboring peaks in the spectrum whose components are due to the different isotopes. Since the detector efficiencies are approximately the same for adjacent peaks, the accuracy of the isotopic ratios is dependent on the half-lives, branching intensities, and measured peak areas. The data presented describe the results obtained by analyzing gamma-ray spectra in the energy region from 120 to 700 keV. Most of the data analyzed were obtained from plutonium material containing 6 percent plutonium-240. Sample weights varied from 0.25 g to approximately 1.2 kg. The methods were also applied to plutonium samples containing up to 23 percent plutonium-240 with weights of 0.25 to 200 g. Results obtained by gamma-ray spectroscopy are compared to chemical analyses of aliquots taken from the bulk samples

  5. Study of the gamma radiation of ionium

    Energy Technology Data Exchange (ETDEWEB)

    Curie, I

    1949-12-01

    A Geiger counter study has been made of the ..gamma.. radiation of ionium. Eleven quanta of the L radiation of radium were observed for every hundred ..cap alpha.. disintegrations, and three ..gamma.. rays were found with energies of 68, 140, and 240 keV at a rate of 0.85, 0.33, 0.05 quanta, respectively, for 100 disintegrations. It is noted that the radiation spectrum of ionium as a whole is difficult to interpret. In the course of this work, the author calculated the efficiency of a thin-walled aluminum counter, both for the L radiation of radium and for ..gamma.. rays of 68 keV. The author also measured, for soft radiation, the ratio between the efficiency of a thin-walled aluminum counter and that of a similar counter lined with 0.11 mm of lead.

  6. Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

    The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron-gamma

  7. The application of gamma-spectrometry to nuclear power plant (NPP) and environment

    International Nuclear Information System (INIS)

    Asgharizadeh, Farid.

    1995-01-01

    One of measuring systems is nuclear spectrometry, particularly Gamma-Ray Spectrometry, to measure and determine the radionuclide concentration within plant materials and environmental samples. There are four major applied techniques related to Nuclear Power Plant operation and environmental monitoring aspects. Some details about gamma ray spectrometry technique is discussed in chapter 2. The main emphasis is on the calculation of gamma-ray detector efficiency for different geometries, the minimum detectable activity concepts and dead-time correction. Also,some formula and relations are introduced. In chapter 3, the major applications of gamma-ray spectrometry for analysis of nuclear power plant and environmental samples are discussed. These applications are divided into four topics: Nuclear Fuel survey; based on the activity of fission products concentration in reactor coolant, two other applications are introduced: Fuel Burnup calculation and the calculation of rated activity of natural radionuclides in construction of materials which is the last and most important application: Measurement and determination of radionuclides activity concentr[[[[n in environmental samples is described through section 3.3 Sampling and measuring methods for research and monitoring aspects is evaluated. Some data about sample preparation methods such as pretreatment and solubilization procedures are presented. Quantitative chemical separations of trace constituents from complex sample materials invariably require meticulous work by an analytical chemist. The radiochemical separation deals with this subject. Instrumental aspects, relate to gamma-ray spectrometry, quality assurance, presentation and reporting of results are described. In the experimental part, determination of radionuclides concentration in sediment sample is presented

  8. External exposure from gamma radiation in uranium mines

    International Nuclear Information System (INIS)

    Thomson, J.E.

    1982-01-01

    Radiation doses received by workers in a high ore grade uranium mine are compared to those of other radiation workers and the need to be able to calculate the exposure rate from an ore body is indicated. The uranium-238 decay chain is presented and particular reference is made to the main gamma emitters and secular equilibrium of the members of the chain. Difficulties in dealing with a self attenuating volume source, in which scattering is important, are pointed out and traditional methods of solution are mentioned. It is shown that in the special case of an infinite ore body a simple solution may be obtained using the energy conservation principle. A straightforward method for calculating the exposure rate from an arbitrarily shaped ore body is given and corrections due to air attenuation, different soil types and possible lack of secular equilibrium are dealt with. The gamma ray spectrum from the ore is discussed with specific reference to the selection of suitable exposure monitors and the calculation of transmission through shields

  9. Gamma-ray multiplicity distribution in ternary fission of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Kliman, J [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Krupa, L [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Morhac, M [Department of Nuclear Physics, Slovak Academy of Sciences, Dubravska cesta 9, Bratislava (Slovakia); Hamilton, J H [Department of Physics, Vanderbilt University, Nashville, TN (United States); Kormicki, J [Department of Physics, Vanderbilt University, Nashville, TN (United States); Ramayya, A V [Department of Physics, Vanderbilt University, Nashville, TN (United States); Hwang, J K [Department of Physics, Vanderbilt University, Nashville, TN (United States); Luo, Y X [Department of Physics, Vanderbilt University, Nashville, TN (United States); Fong, D [Department of Physics, Vanderbilt University, Nashville, TN (United States); Gore, P [Department of Physics, Vanderbilt University, Nashville, TN (United States); Akopian, G M Ter; Oganessian, Yu Ts; Rodin, A M; Fomichev, A S; Popeko, G S; Daniel, A V [Flerov Laboratory for Nuclear Reactions, Joint Institute for Nuclear Research, Dubna (Russian Federation); Rasmussen, J O; Macchiavelli, A O [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Stoyer, M A [Lawrence Livermore National Laboratory, Livermore, CA (United States); Donangelo, R [Instituto de Fisica, Universidade Federal do Rio de Janeiro, 21945-970 Rio de Janeiro (Brazil); Cole, J D [Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID (United States)

    2002-12-01

    From multiparameter data obtained at Lawrence Berkeley National Laboratory, the integral characteristics of the prompt {gamma}-ray emission were extracted for tripartition of {sup 252}Cf with He, Be and C being the third light charged particle. We used multifold {gamma}-ray coincidence spectra for the determination of {gamma}-ray multiplicities assuming a Gaussian distribution for {gamma}-ray multiplicity. The multiplicity distribution characteristics, i.e. mean multiplicity and its dispersion were obtained by minimizing with respect to the calculated values of probabilities of multifold {gamma}-ray coincidences using a combinatoric method. Comparison with the known experimental data from binary fission was made. Further, we investigated dependencies of the mean {gamma}-ray multiplicity on the kinetic energy of the light charged particle. The mean {gamma}-ray multiplicity for He ternary fission is found to increase rapidly with increasing kinetic energy of He in the region less than 11 MeV and then decrease slowly with increasing kinetic energy of He. The anomalous behaviour of {gamma}-ray emission is discussed. The mean {gamma}-ray multiplicity was determined for the first time for Be and C ternary fission. For Be, the {gamma}-ray multiplicity as a function of kinetic energy was obtained as well.

  10. Graphical comparison of calculated internal conversion coefficients

    International Nuclear Information System (INIS)

    Ewbank, W.B.

    1980-11-01

    Calculated values of the coefficients of internal conversion of gamma rays in the K shell and L 1 , L 2 , L 3 subshells from published tabulations by Band and Trzhaskovskaya and by Roesel et al. at Data Nucl. Data Tables, 21, 92-514(1978) are compared with values obtained by computer interpolation among tabulated values of Hager and Seltzer Nucl. Data, A4, 1-235(1968). In some cases, agreement among the three calculations is remarkably good, and differences are generally less than 5%. In a few cases, there are differences as large as 20 to 50%, corresponding to the threshold effect described by Roesel et al. The Z-dependent resonance minimum described by Roesel et al. is also observed in the comparison of E1-E4 conversion in the L 1 subshell. In several cases (notably M1-M4 conversion in the K shell and L 1 subshell), the Band and Roesel calculations show dramatically different dependence on gamma energy and atomic number. For Z = 100, the Band calculation for E4 conversion in the L 3 subshell shows irregular behavior at energies below the K-shell binding energy. A few high-quality measurements of internal conversion coefficients (+-5%) would help greatly to establish a basis for choice among the theoretical calculations. 32 figures

  11. SWEPP gamma-ray spectrometer system software user's guide

    International Nuclear Information System (INIS)

    Femec, D.A.

    1994-08-01

    The SWEPP Gamma-Ray Spectrometer (SGRS) System has been developed by the Radiation Measurement and Development Unit of the Idaho National Engineering Laboratory to assist in the characterization of the radiological contents of contact-handled waste containers at the Stored Waste Examination Pilot Plant (SWEPP). In addition to determining the concentrations of gamma-ray-emitting radionuclides, the software also calculates attenuation-corrected isotopic mass ratios of specific interest, and provides controls for SGRS hardware as required. This document serves as a user's guide for the data acquisition and analysis software associated with the SGRS system

  12. Gamma ray spectrum analysis code: sigmas 1.0

    International Nuclear Information System (INIS)

    Siangsanan, P.; Dharmavanij, W.; Chongkum, S.

    1996-01-01

    We have developed Sigmas 1.0 a software package for data reduction and gamma ray spectra evaluation. It is capable of analysing the gamma-ray spectrum in the range of 0-3 MeV by semiconductor detector, i.e. Ge(Li) or HPGe, peak searching, net area determining, plotting and spectrum displaying. There are two methods for calculating the net area under peaks; the Covell method and non-linear fitting by the method of Levenberg and Marquardt which can fit any multiplet peak in the spectrum. The graphic display was rather fast and user friendly

  13. Application of the similitude principle to gamma-gamma density measurements; Application du principe de similitude a la mesure gamma-gamma de densite

    Energy Technology Data Exchange (ETDEWEB)

    Czubek, J A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires. Departement d' Electronique Generale, Service d' Electronique Industrielle; Institut de Recherches Nucleaires, Dep. VI, Cracow (Poland)

    1966-07-01

    The work presented here deals with the problem of the application of the similitude principle to rock density measurements by the gamma-gamma method. A formula is presented which makes it possible to transform results of gamma-gamma measurements carried out on models in order to make them suitable for comparison with results obtained under actual field conditions. Both the space coordinates and the densities are transformed. This transformation makes it possible to obtain a calibration curve as a function of the density for a gamma-gamma probe using only a single model of given density. The influence has also been studied of the chemical composition on the results obtained from gamma-gamma measurements. A method has been developed for estimating the equivalent Z parameter of the medium; the possibility of completely eliminating the influence of the chemical composition of the medium on the measurement results has been studied. (author) [French] L'etude presentee ci-dessous traite le probleme de l'application du principe de similitude aux mesures de densite des roches par la methode gamma-gamma. Nous indiquons une formule qui permet de transformer les resultats des mesures gamma-gamma effectuees sur les modeles pour les comparer aux resultats obtenus dans les conditions reelles sur le terrain. On transforme les coordonnees spatiales ainsi que les densites. Cette transformation donne la possibilite d'obtenir une courbe d'etalonnage (en fonction de la densite) pour une sonde gamma-gamma en utilisant un seul modele de densite donnee. On a etudie aussi l'influence de la composition chimique sur les resultats obtenus des mesures gamma-gamma. On a etabli une methode d'estimation du parametre Z equivalent du milieu, ainsi que la possibilite d'eliminer completement l'influence de la composition chimique du milieu sur les resultats des mesures de densite. (auteur)

  14. {gamma} activity and heating of rods in EL2 and EL3; Activitiy {gamma} et echauffement des barres de EL2 et EL3

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    A method is described for calculating the {gamma} activity of uranium rods, given the mean flux in which they are irradiated, the time they remain in the pile and the duration of deactivation. This calculation leads to numerical formulae which may be applied to the rods of the two reactors. It allows the saturation activities to be foreseen both for EL2 and for EL3, taking into recount the minimum times necessary for extraction. Measurements have been carried out, and the results are in good agreement with those foreseen by calculation. In the last section this method is used to calculate the heating of the irradiated rods. (author) [French] Une methode est indiquee ici, qui permet de calculer l'activite {gamma} des barres d'uranium connaissant le flux moyen dans lequel elles ont ete irradiees, leur temps de sejour en pile et la duree de la desactivation. Ce calcul conduit a des formules numeriques que l'on peut appliquer aux barres des deux reacteurs. Il permet de prevoir les activites atteintes a saturation, tant a EL2 qu'a EL3, compte tenu des temps minima necessaires a l'extraction. Des mesures ont ete faites: les resultats sont en bon accord avec les previsions du calcul. Enfin, en derniere partie, cette methode est utilisee pour calculer l'echauffement des barres irradiees. (auteur)

  15. Evaluation of gamma dose effect on PIN photodiode using analytical model

    Science.gov (United States)

    Jafari, H.; Feghhi, S. A. H.; Boorboor, S.

    2018-03-01

    The PIN silicon photodiodes are widely used in the applications which may be found in radiation environment such as space mission, medical imaging and non-destructive testing. Radiation-induced damage in these devices causes to degrade the photodiode parameters. In this work, we have used new approach to evaluate gamma dose effects on a commercial PIN photodiode (BPX65) based on an analytical model. In this approach, the NIEL parameter has been calculated for gamma rays from a 60Co source by GEANT4. The radiation damage mechanisms have been considered by solving numerically the Poisson and continuity equations with the appropriate boundary conditions, parameters and physical models. Defects caused by radiation in silicon have been formulated in terms of the damage coefficient for the minority carriers' lifetime. The gamma induced degradation parameters of the silicon PIN photodiode have been analyzed in detail and the results were compared with experimental measurements and as well as the results of ATLAS semiconductor simulator to verify and parameterize the analytical model calculations. The results showed reasonable agreement between them for BPX65 silicon photodiode irradiated by 60Co gamma source at total doses up to 5 kGy under different reverse voltages.

  16. Gamma-gamma directional correlations and coincidence studies in 154Gd

    International Nuclear Information System (INIS)

    Gupta, J.B.; Gupta, S.L.; Hamilton, J.H.; Ramayya, A.V.; Delhi Univ.

    1977-01-01

    The intensities, placements and E2/M1 mixing ratios of transitions in the decay of 154 Eu have been carefully studied to provide accurate data for microscopic calculations. Coincidence relationships in thhe decay of 154 Eu have been studied extensively with a multiparameter γ-γ coincidence system with two large volume Ge(Li) detectors. Spectra in coincidence with twenty energy gates were analyzed. Twenty-nine new coincidence relationships were established and confirmed most, but not all, of several levels previously assigned by energy fits only. From an analysis of coincidence spectra and singles spectra with a 18% efficiency Ge(Li) detector new information on the gamma-ray intensities were obtained. Precise values of the E2/M1 mixing ratios of transitions from the gamma- and beta-vibrational bands to the g.s. band have been determined from γ-γ directional correlation measurements with a NaI(Tl)-Ge(Li) detector coincidence system. Mixing ratios were obtained for a number of other transitions including those from KPI = 0 - and 2+ bands from direct and skipped cascade correlations. (orig.) [de

  17. RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

    International Nuclear Information System (INIS)

    Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1977-07-01

    The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

  18. Research on 3-D terrain correction methods of airborne gamma-ray spectrometry survey

    International Nuclear Information System (INIS)

    Liu Yanyang; Liu Qingcheng; Zhang Zhiyong

    2008-01-01

    The general method of height correction is not effectual in complex terrain during the process of explaining airborne gamma-ray spectrometry data, and the 2-D terrain correction method researched in recent years is just available for correction of section measured. A new method of 3-D sector terrain correction is studied. The ground radiator is divided into many small sector radiators by the method, then the irradiation rate is calculated in certain survey distance, and the total value of all small radiate sources is regarded as the irradiation rate of the ground radiator at certain point of aero- survey, and the correction coefficients of every point are calculated which then applied to correct to airborne gamma-ray spectrometry data. The method can achieve the forward calculation, inversion calculation and terrain correction for airborne gamma-ray spectrometry survey in complex topography by dividing the ground radiator into many small sectors. Other factors are considered such as the un- saturated degree of measure scope, uneven-radiator content on ground, and so on. The results of for- ward model and an example analysis show that the 3-D terrain correction method is proper and effectual. (authors)

  19. The development of gamma energy identify algorithm for compact radiation sensors using stepwise refinement technique

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hyun Jun [Div. of Radiation Regulation, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Ye Won; Kim, Hyun Duk; Cho, Gyu Seong [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Yi, Yun [Dept. of of Electronics and Information Engineering, Korea University, Seoul (Korea, Republic of)

    2017-06-15

    A gamma energy identifying algorithm using spectral decomposition combined with smoothing method was suggested to confirm the existence of the artificial radio isotopes. The algorithm is composed by original pattern recognition method and smoothing method to enhance the performance to identify gamma energy of radiation sensors that have low energy resolution. The gamma energy identifying algorithm for the compact radiation sensor is a three-step of refinement process. Firstly, the magnitude set is calculated by the original spectral decomposition. Secondly, the magnitude of modeling error in the magnitude set is reduced by the smoothing method. Thirdly, the expected gamma energy is finally decided based on the enhanced magnitude set as a result of the spectral decomposition with the smoothing method. The algorithm was optimized for the designed radiation sensor composed of a CsI (Tl) scintillator and a silicon pin diode. The two performance parameters used to estimate the algorithm are the accuracy of expected gamma energy and the number of repeated calculations. The original gamma energy was accurately identified with the single energy of gamma radiation by adapting this modeling error reduction method. Also the average error decreased by half with the multi energies of gamma radiation in comparison to the original spectral decomposition. In addition, the number of repeated calculations also decreased by half even in low fluence conditions under 104 (/0.09 cm{sup 2} of the scintillator surface). Through the development of this algorithm, we have confirmed the possibility of developing a product that can identify artificial radionuclides nearby using inexpensive radiation sensors that are easy to use by the public. Therefore, it can contribute to reduce the anxiety of the public exposure by determining the presence of artificial radionuclides in the vicinity.

  20. The induced radioactivity danger parameter for gamma radiation

    International Nuclear Information System (INIS)

    Perry, D.R.

    1985-07-01

    Dosimetric and practical aspects of the induced radioactivity danger parameter, as used for calculating the gamma radiation dose rate near to objects that have been exposed to high energy radiation, are examined. A simplified and more generally applicable method of calculation is proposed, based on energy balance in homogeneous media. The problems of applying this in practice are discussed, and it is shown that corrections are generally small enough to be neglected in many practical applications. Examples of calculations by previous and proposed methods are given. (author)

  1. Calculation of the counting efficiency for extended sources

    International Nuclear Information System (INIS)

    Korun, M.; Vidmar, T.

    2002-01-01

    A computer program for calculation of efficiency calibration curves for extended samples counted on gamma- and X ray spectrometers is described. The program calculates efficiency calibration curves for homogeneous cylindrical samples placed coaxially with the symmetry axis of the detector. The method of calculation is based on integration over the sample volume of the efficiencies for point sources measured in free space on an equidistant grid of points. The attenuation of photons within the sample is taken into account using the self-attenuation function calculated with a two-dimensional detector model. (author)

  2. Nuclear power development and nuclear data activities in Malaysia

    International Nuclear Information System (INIS)

    Gui Ah Auu

    1999-01-01

    In this paper, research activities on nuclear power requirement carried out jointly by MINT and other organizations are described. Also discussed are activities on neutronics such as TRIGA reactor fuel management, storage pool criticality, and reactor fuel transfer cask calculations. In addition, recent work on radiation transport activities in MINT such as skyshine and photon phantom dose calculations using the MCNP and MRIPP computer codes are presented. Finally, nuclear data measurement works by researchers in Malaysian universities are described. (author)

  3. Nuclear power development and nuclear data activities in Malaysia

    Energy Technology Data Exchange (ETDEWEB)

    Gui Ah Auu [Malaysian Institute for Nuclear Technology Research, Ministry of Science, Technology and the Environment, Selangor (Malaysia)

    1999-03-01

    In this paper, research activities on nuclear power requirement carried out jointly by MINT and other organizations are described. Also discussed are activities on neutronics such as TRIGA reactor fuel management, storage pool criticality, and reactor fuel transfer cask calculations. In addition, recent work on radiation transport activities in MINT such as skyshine and photon phantom dose calculations using the MCNP and MRIPP computer codes are presented. Finally, nuclear data measurement works by researchers in Malaysian universities are described. (author)

  4. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  5. Simulation Study on Identifiability of UHE Gamma-ray Air Showers

    International Nuclear Information System (INIS)

    Wada, Y.; Inoue, N.; Miyazawa, K.; Vankov, H.P.

    2008-01-01

    The chemical composition of Ultra-High-Energy (UHE) comic rays is one of unsolved mysteries, and its study will give us fruitful information on the origin and acceleration mechanism of UHE cosmic rays. Especially, a detection of UHE gamma-rays by hybrid experiments, such as AUGER and TA, will be a key to solve these questions. The characteristics of UHE gamma-ray showers have been studied by comparing the lateral and longitudinal structures of shower particles calculated with AIRES and our own simulation code, so far. There are apparent differences in a slope of lateral distribution (η) and a depth of shower maximum (Xmax) between gamma-ray and proton induced showers because UHE gamma-ray showers are affected by the LPM effect and the geomagnetic cascading process in an energy region of >10 19.5 eV. Different features between gamma-ray and proton showers are pointed out from the simulation study and an identifiability of gamma-ray showers from proton ones is also discussed by the method of Neural-Network-Analysis

  6. Simulation Study on Identifiability of UHE Gamma-ray Air Showers

    Energy Technology Data Exchange (ETDEWEB)

    Wada, Y.; Inoue, N.; Miyazawa, K. [Graduate School of Science and Engineering, Saitama University, Saitama 338-8570 (Japan); Vankov, H.P. [Institute for Nuclear Research and Nuclear Energy, Bulgaria Academy, Sofia (Bulgaria)

    2008-01-15

    The chemical composition of Ultra-High-Energy (UHE) comic rays is one of unsolved mysteries, and its study will give us fruitful information on the origin and acceleration mechanism of UHE cosmic rays. Especially, a detection of UHE gamma-rays by hybrid experiments, such as AUGER and TA, will be a key to solve these questions. The characteristics of UHE gamma-ray showers have been studied by comparing the lateral and longitudinal structures of shower particles calculated with AIRES and our own simulation code, so far. There are apparent differences in a slope of lateral distribution ({eta}) and a depth of shower maximum (Xmax) between gamma-ray and proton induced showers because UHE gamma-ray showers are affected by the LPM effect and the geomagnetic cascading process in an energy region of >10{sup 19.5}eV. Different features between gamma-ray and proton showers are pointed out from the simulation study and an identifiability of gamma-ray showers from proton ones is also discussed by the method of Neural-Network-Analysis.

  7. Contributions to indoor gamma dose rate from building materials

    International Nuclear Information System (INIS)

    Liu Xionghua; Li Guangming; Yang Xiangdong

    1990-01-01

    In the coures of construction of a building structured with bricks and concrets, the indoor gamma air absorbed dose rates were seperately measured from the floors, brick walls and prefabricated plates of concrets, etc.. It suggested that the indoor gamma dose rates from building materials are mainly attributed to the brick walls and the floors. A little contribution comes from other brilding materials. The dose rates can be calculated through a 4π-infinite thick model with a correction factor of 0.52

  8. Spreadsheet analysis of gamma spectra for nuclear material measurements

    International Nuclear Information System (INIS)

    Mosby, W.R.; Pace, D.M.

    1990-01-01

    A widely available commercial spreadsheet package for personal computers is used to calculate gamma spectra peak areas using both region of interest and peak fitting methods. The gamma peak areas obtained are used for uranium enrichment assays and for isotopic analyses of mixtures of transuranics. The use of spreadsheet software with an internal processing language allows automation of routine analysis procedures increasing ease of use and reducing processing errors while providing great flexibility in addressing unusual measurement problems. 4 refs., 9 figs

  9. Impaired theta-gamma coupling during working memory performance in schizophrenia.

    Science.gov (United States)

    Barr, Mera S; Rajji, Tarek K; Zomorrodi, Reza; Radhu, Natasha; George, Tony P; Blumberger, Daniel M; Daskalakis, Zafiris J

    2017-11-01

    Working memory deficits represent a core feature of schizophrenia. These deficits have been associated with dysfunctional dorsolateral prefrontal cortex (DLPFC) cortical oscillations. Theta-gamma coupling describes the modulation of gamma oscillations by theta phasic activity that has been directly associated with the ordering of information during working memory performance. Evaluating theta-gamma coupling may provide greater insight into the neural mechanisms mediating working memory deficits in this disorder. Thirty-eight patients diagnosed with schizophrenia or schizoaffective disorder and 38 healthy controls performed the verbal N-Back task administered at 4 levels, while EEG was recorded. Theta (4-7Hz)-gamma (30-50Hz) coupling was calculated for target and non-target correct trials for each working memory load. The relationship between theta-gamma coupling and accuracy was determined. Theta-gamma coupling was significantly and selectively impaired during correct responses to target letters among schizophrenia patients compared to healthy controls. A significant and positive relationship was found between theta-gamma coupling and 3-Back accuracy in controls, while this relationship was not observed in patients. These findings suggest that impaired theta-gamma coupling contribute to working memory dysfunction in schizophrenia. Future work is needed to evaluate the predictive utility of theta-gamma coupling as a neurophysiological marker for functional outcomes in this disorder. Copyright © 2017. Published by Elsevier B.V.

  10. Nature of gamma rays background radiation in new and old buildings of Qatar University

    International Nuclear Information System (INIS)

    Al-Houty, L.; Abou-Leila, H.; El-Kameesy, S.

    1987-01-01

    Measurements and analysis of gamma-background radiation spectrum in four different places of Qatar University campus were performed at the energy range 10 keV-3 MeV using hyper pure Ge-detector. The dependence of the detector absolute photopeak efficiency on gamma-ray energies was determined and correction of the data for that was also done. The absorbed dose for each gamma line was calculated and an estimation of the total absorbed dose for the detected gamma lines in the four different places was obtained. Comparison with other results was also performed

  11. Testing of massive lead containers by gamma densitometry

    International Nuclear Information System (INIS)

    Janardhanan, S.; Dabhadkar, S.B.; Subbaratnam, T.

    1977-01-01

    A non-destructive method of testing the shielding adequacy of transport and hold-up containers for radioactive sources and waste is described. The method involves measurement of the gamma intensity transmitted through the shield by a radioactive gamma source located inside. The data obtained is used to correlate the intensity with the lead thickness and thereby detect, locate and assess the extent of damage or faults if any so that corrective action can be taken in time. Factors influencing the choice of the gamma source, its strength and means of detection are described. Methods of checking the results of measurement with calculated values are outlined. The advantages of the method, its reliability and expediency with which the method can be adopted to varying applications make it an unique application in reactor and isotopes technology. (author)

  12. Study of {gamma} radiation from uranium rods during deactivation; Etude du rayonnement {gamma} des barres d'uranium en court de desactivation

    Energy Technology Data Exchange (ETDEWEB)

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

    1957-07-01

    The classical formulae giving the {gamma} activities of the fission products contained in a uranium rod after unloading from the pile are reviewed without being proved. The knowledge of these activities makes it possible, by means of the method proposed here, to determine the intensities of ionisation at a point outside the rod, and thus to establish {gamma} radiation diagrams. The different parameters introduced in the calculation are geometric (dimensions of the bars and coordinates of the point considered), energetic (power at which the bar has been irradiated) and temporal (duration of the irradiation and deactivation). A numerical example follows the demonstration of the general formulae, {gamma} flux measurements carried out in the deactivation well of P2 (Saclay pile) define the accuracy of the method. In conclusion, it is suggested that radiation diagrams be used in (planning the use of) industrial irradiators for radiochemical polymerisation or the preservation of food products. (author) [French] On rappelle sans demonstration les formules classiques donnant les activites {gamma} des produits de fission contenus dans une barre d'uranium apres defournement. La connaissance de ces activites permet par la methode proposee de passer aux intensites d'ionisation en un point exterieur a la barre et d'etablir ainsi des diagrammes de rayonnement {gamma}. Les differents parametres introduits dans le calcul sont d'ordre geometrique (dimensions des barres et coordonnees du point considere), d'ordre energetique (puissance a laquelle la barre a ete irradiee) et fonction du temps (duree d'irradiation et de desactivation). Un exemple numerique fait suite a la demonstration des formules generales. Des mesures de flux {gamma} effectuees au puits de desactivation de P2 (pile de Saclay) fixent le degre d'approximation de la methode. En conclusion, on suggere l'utilisation des diagrammes de rayonnement dans l'etablissement de projets d'irradiateurs industriels pour les

  13. Utilization of ilmenite/epoxy composite for neutrons and gamma rays attenuation

    Energy Technology Data Exchange (ETDEWEB)

    El-Sayed Abdo, A. E-mail: attiaabdo11@hotmail.com; El-Sarraf, M.A.; Gaber, F.A

    2003-01-01

    This work deals with the study of ilmenite/epoxy composite as an injecting mortar for cracks developed in biological concrete shields, as well as, neutrons and gamma rays attenuation. Effects of the particle size on the mechanical strengths have been studied for epoxy resin filled with crushed ilmenite with different maximum particle sizes ranging from 32 to 500 {mu}m. Thermal neutrons and gamma rays attenuation in ilmenite/epoxy composites with 75 and 80 wt.% of ilmenite concentration have been investigated. The total mass attenuation coefficients {mu}/{rho} (cm{sup 2} g{sup -1}) of gamma ray for five ilmenite/epoxy composites have been calculated using the XCOM program (version 3.1) at energies from 10 keV to 100 MeV. Also, the total mass attenuation coefficients ({mu}/{rho}) have estimated based on the measured total linear attenuation coefficients ({mu}) and compared with the calculated results where, a reasonable agreement was found.

  14. Public effective doses from environmental natural gamma exposures indoors and outdoors in Iran

    International Nuclear Information System (INIS)

    Sohrabi, Mehdi; Roositalab, Jalil; Mohammadi, Jahangir

    2015-01-01

    The effective doses of public in Iran due to external gamma exposures from terrestrial radionuclides and from cosmic radiation indoors and outdoors of normal natural background radiation areas were determined by measurements and by calculations. For direct measurements, three measurement methods were used including a NaI(TI) scintillation survey meter for preliminary screening, a pressurised ionising chamber for more precise measurements and early warning measurement equipment systems. Measurements were carried out in a large number of locations indoors and outdoors ∼1000 houses selected randomly in 36 large cities of Iran. The external gamma doses of public from living indoors and outdoors were also calculated based on the radioactivity measurements of samples taken from soil and building materials by gamma spectrometry using a high-resolution HPGe system. The national mean background gamma dose rates in air indoors and outdoors based on measurements are 126.9±24.3 and 111.7±17.72 nGy h -1 , respectively. When the contribution from cosmic rays was excluded, the values indoors and outdoors are 109.2±20.2 and 70.2±20.59.4 nGy h -1 , respectively. The dose rates determined for indoors and outdoors by calculations are 101.5±9.2 and 72.2±9.4 nGy h -1 , respectively, which are in good agreement with directly measured dose rates within statistical variations. By considering a population-weighted mean for terrestrial radiation, the ratio of indoor to outdoor dose rates is 1.55. The mean annual effective dose of each individual member of the public from terrestrial radionuclides and cosmic radiation, indoors and outdoors, is 0.86±0.16 mSv y -1 by measurements and 0.8±0.2 mSv y -1 by calculations. The results of this national survey of public annual effective doses from national natural background external gamma radiation determined by measurements and calculations indoors and outdoors of 1000 houses in 36 cities of Iran are presented and discussed. (authors)

  15. High resolution {sup 12}C({gamma},p) experiments at E{sub {gamma}} {approx_equal} 25-75 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Ruijter, H

    1995-08-01

    Absolute differential cross sections for the {sup 12}C({gamma},p){sup 11}B reaction have been measured over proton detection angels ranging from 30 to 150 deg, using tagged photons of 25-75 MeV energy, for low-lying regions of residual excitation energy in {sup 11}B. Four experiments were performed at the MAX laboratory in Lund in order to provide data. Previously reported cross sections for the reaction had systematic uncertainties of a magnitude which made them agree, in spite of a large spread in absolute values. The cross sections reported, with a systematic uncertainty of 8%, remove previous ambiguities for E{sub {gamma}}=40-75 MeV. A reinterpretation of the states excited in{sup 11}B at E about 7 MeV is also presented. The data are compared with quasi-elastic (e,e`p) results in PWIA in the same recoil momentum range. It is found that the momentum distributions do not scale for the two reaction types. Furthermore, the data are compared with the results for the inverse reaction (p,{gamma}) in the centre-of-momentum system by detailed balance. The comparison with respect to missing momentum indicates an angular dependence in the ({gamma},p) reaction which is not present in the inverse (p,{gamma}) reaction. Recent results from the MAX laboratory for the ({gamma},n) reaction are compared to the ({gamma},p) results. The mirror nuclei {sup 11}C and {sup 11}B have almost identical excitation energy spectra at E{sub {gamma}}=60 MeV. It is concluded that HF-RPA calculations with essential contributions of meson exchange currents provide a qualitative description of the angular distributions obtained for the ({gamma},p) reaction. An extension of the spherical symmetric basis for the wave function is suggested for the states at E about 7 MeV in {sup 11}B. 108 refs, 83 figs.

  16. Monte Carlo calculation of the total probability for gamma-Ray interaction in toluene

    International Nuclear Information System (INIS)

    Grau Malonda, A.; Garcia-Torano, E.

    1983-01-01

    Interaction and absorption probabilities for gamma-rays with energies between 1 and 1000 KeV have been computed and tabulated. Toluene based scintillator solution has been assumed in the computation. Both, point sources and homogeneously dispersed radioactive material have been assumed. These tables may be applied to cylinders with radii between 1.25 cm and 0.25 cm and heights between 4.07 cm and 0.20 cm. (Author) 26 refs

  17. Two‐year experience with the commercial Gamma Knife Check software

    Science.gov (United States)

    Bhatnagar, Jagdish; Bednarz, Greg; Novotny, Josef; Flickinger, John; Lunsford, L. Dade; Huq, M. Saiful

    2016-01-01

    The Gamma Knife Check software is an FDA approved second check system for dose calculations in Gamma Knife radiosurgery. The purpose of this study was to evaluate the accuracy and the stability of the commercial software package as a tool for independent dose verification. The Gamma Knife Check software version 8.4 was commissioned for a Leksell Gamma Knife Perfexion and a 4C unit at the University of Pittsburgh Medical Center in May 2012. Independent dose verifications were performed using this software for 319 radiosurgery cases on the Perfexion and 283 radiosurgery cases on the 4C units. The cases on each machine were divided into groups according to their diagnoses, and an averaged absolute percent dose difference for each group was calculated. The percentage dose difference for each treatment target was obtained as the relative difference between the Gamma Knife Check dose and the dose from the tissue maximum ratio algorithm (TMR 10) from the GammaPlan software version 10 at the reference point. For treatment plans with imaging skull definition, results obtained from the Gamma Knife Check software using the measurement‐based skull definition method are used for comparison. The collected dose difference data were also analyzed in terms of the distance from the treatment target to the skull, the number of treatment shots used for the target, and the gamma angles of the treatment shots. The averaged percent dose differences between the Gamma Knife Check software and the GammaPlan treatment planning system are 0.3%, 0.89%, 1.24%, 1.09%, 0.83%, 0.55%, 0.33%, and 1.49% for the trigeminal neuralgia, acoustic neuroma, arteriovenous malformation (AVM), meningioma, pituitary adenoma, glioma, functional disorders, and metastasis cases on the Perfexion unit. The corresponding averaged percent dose differences for the 4C unit are 0.33%, 1.2%, 2.78% 1.99%, 1.4%, 1.92%, 0.62%, and 1.51%, respectively. The dose difference is, in general, larger for treatment targets in the

  18. Analysis of gamma irradiated pepper constituents, 5

    International Nuclear Information System (INIS)

    Takagi, Kazuko; Okuyama, Tsuneo; Ishikawa, Toshihiro.

    1988-01-01

    Gamma irradiated peppers (10 krad, 100 krad, 1 Mrad) were analyzed by HPLC. The extraction method and HPLC conditions were same as the first report, that is, the extraction from pepper was performed by Automatic Air Hammer and the extracted samples were separated on a reversed phase C 8 column with a concave gradient from 0.1% trifluoro aceticacid (TFA) in water to 75% acetonitrile-0.1% TFA in water for 60 minutes and detected at 210 nm, 280 nm. It is difficult to compare with irradiated and unirradiated pepper constituents by their peak height or area. And the method of multi variant statistically analysis was introduced. The 'peak n area/peak n + 1 area' ratio was calculated by computer. Each peak area was accounted by integrator. The value of these ratio were called 'parameter'. Each chromatogram has 741 parameters calculated with 39 chromatographic peaks. And these parameters were abopted to the multi variant statiscally analysis. Comparison of constituents between irradiated pepper and unirradiated pepper was done by 741 parameters. The correlation of parameters between irradiated and unirradiated was investigated by use of computer. Some parameters of irradiated case were selected as which had no correlation with unirradiated case. That is to say these parameters were thought to be changed with gamma spectrum irradiation. By this method, Coumarin was identified as a changed component with gamma irradiation. (author)

  19. A novel method for quantitative geosteering using azimuthal gamma-ray logging.

    Science.gov (United States)

    Yuan, Chao; Zhou, Cancan; Zhang, Feng; Hu, Song; Li, Chaoliu

    2015-02-01

    A novel method for quantitative geosteering by using azimuthal gamma-ray logging is proposed. Real-time up and bottom gamma-ray logs when a logging tool travels through a boundary surface with different relative dip angles are simulated with the Monte Carlo method. Study results show that response points of up and bottom gamma-ray logs when the logging tool moves towards a highly radioactive formation can be used to predict the relative dip angle, and then the distance from the drilling bit to the boundary surface is calculated. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. Radiation-induced reactions of Cl-, CO32-, and Br- in seawater, - Model calculation of gamma radiolysis of seawater

    International Nuclear Information System (INIS)

    Hata, Kuniki; Hanawa, Satoshi; Kasahara, Shigeki; Muroya, Yusa; Katsumura, Yosuke

    2012-09-01

    Gamma-radiolysis of seawater has been simulated to estimate the concentrations of radiolysis products. Although gas products such as H 2 , O 2 and H 2 O 2 in irradiated pure water quickly attain the steady state with very low concentrations, the products in seawater monotonically increase with dose. It was found that H 2 is produced almost linearly with dose, and corresponding G-value was 4.4 x 10 -8 mol J -1 . As similar result was obtained from the calculation of 8 x 10 -4 mol dm -3 NaBr solution, the origin of the linear increase in seawater was attributable to be the reactions of Br - . According to the sensitivity analysis, three reactions, 1: Br - + ·OH → BrOH· - , 2: BrOH· - → Br - + ·OH, and 3: BrOH· - → Br· + OH - , determined the concentrations of the products. The presence of Cl - and HCO 3 - in seawater hardly affected the concentrations of the radiolysis products. Oxyanions derived from Cl - and Br - were not obtained at observable concentration. (authors)

  1. Approximate calculational techniques for radiation protection applications (collection of papers presented at the November 1985 American Nuclear Society meeting)

    Energy Technology Data Exchange (ETDEWEB)

    Rice, A.F.; Roussin, R.W. (comps.)

    1986-09-01

    Although radiation protection principles are, on the whole, well understood and a whole series of computer codes exist for their solution, it is felt that there is a need for practical, approximate techniques to be used by the practicing nuclear engineer for a variety of applications. Within the context of approximate techniques, the papers presented cover a broad overview of specific problems, for example, skyshine and penetration analysis, with applications extending from general nuclear reactor design to spent fuel storage and fusion. Separate abstracts have been prepared for individual papers.

  2. Failure modes and effects analysis (FMEA) for Gamma Knife radiosurgery.

    Science.gov (United States)

    Xu, Andy Yuanguang; Bhatnagar, Jagdish; Bednarz, Greg; Flickinger, John; Arai, Yoshio; Vacsulka, Jonet; Feng, Wenzheng; Monaco, Edward; Niranjan, Ajay; Lunsford, L Dade; Huq, M Saiful

    2017-11-01

    Gamma Knife radiosurgery is a highly precise and accurate treatment technique for treating brain diseases with low risk of serious error that nevertheless could potentially be reduced. We applied the AAPM Task Group 100 recommended failure modes and effects analysis (FMEA) tool to develop a risk-based quality management program for Gamma Knife radiosurgery. A team consisting of medical physicists, radiation oncologists, neurosurgeons, radiation safety officers, nurses, operating room technologists, and schedulers at our institution and an external physicist expert on Gamma Knife was formed for the FMEA study. A process tree and a failure mode table were created for the Gamma Knife radiosurgery procedures using the Leksell Gamma Knife Perfexion and 4C units. Three scores for the probability of occurrence (O), the severity (S), and the probability of no detection for failure mode (D) were assigned to each failure mode by 8 professionals on a scale from 1 to 10. An overall risk priority number (RPN) for each failure mode was then calculated from the averaged O, S, and D scores. The coefficient of variation for each O, S, or D score was also calculated. The failure modes identified were prioritized in terms of both the RPN scores and the severity scores. The established process tree for Gamma Knife radiosurgery consists of 10 subprocesses and 53 steps, including a subprocess for frame placement and 11 steps that are directly related to the frame-based nature of the Gamma Knife radiosurgery. Out of the 86 failure modes identified, 40 Gamma Knife specific failure modes were caused by the potential for inappropriate use of the radiosurgery head frame, the imaging fiducial boxes, the Gamma Knife helmets and plugs, the skull definition tools as well as other features of the GammaPlan treatment planning system. The other 46 failure modes are associated with the registration, imaging, image transfer, contouring processes that are common for all external beam radiation therapy

  3. The Development of Gamma Irradiator Control System

    International Nuclear Information System (INIS)

    Mohd Zaid Hassan; Anwar Abdul Rahman; Azraf Azman; Mohd Rizal Mamat

    2015-01-01

    This paper presents the preliminary software development for the Gamma irradiator control system using commercial supervisory control and data acquisition (SCADA) software. The radiation dose analysis is the study of the relationship between the initial loading source activity (Curie) and concurrent activity in order to perform the irradiation process. The concurrent source activity calculation model is presented. The Human machine interface (HMI) has been developed by using Indusoft Web Studio to solve the mathematical calculation, task and process overview. (author)

  4. Neutron/gamma dose separation by the multiple-ion-chamber technique

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1983-01-01

    Many mixed n/γ dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and 60 Co gamma ray irradiation in air. The MORSE-CG coupled n/γ three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references

  5. OGRE, Monte-Carlo System for Gamma Transport Problems

    International Nuclear Information System (INIS)

    1984-01-01

    1 - Nature of physical problem solved: The OGRE programme system was designed to calculate, by Monte Carlo methods, any quantity related to gamma-ray transport. The system is represented by two examples - OGRE-P1 and OGRE-G. The OGRE-P1 programme is a simple prototype which calculates dose rate on one side of a slab due to a plane source on the other side. The OGRE-G programme, a prototype of a programme utilizing a general-geometry routine, calculates dose rate at arbitrary points. A very general source description in OGRE-G may be employed by reading a tape prepared by the user. 2 - Method of solution: Case histories of gamma rays in the prescribed geometry are generated and analyzed to produce averages of any desired quantity which, in the case of the prototypes, are gamma-ray dose rates. The system is designed to achieve generality by ease of modification. No importance sampling is built into the prototypes, a very general geometry subroutine permits the treatment of complicated geometries. This is essentially the same routine used in the O5R neutron transport system. Boundaries may be either planes or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. Cross section data is prepared by the auxiliary master cross section programme XSECT which may be used to originate, update, or edit the master cross section tape. The master cross section tape is utilized in the OGRE programmes to produce detailed tables of macroscopic cross sections which are used during the Monte Carlo calculations. 3 - Restrictions on the complexity of the problem: Maximum cross-section array information may be estimated by a given formula for a specific problem. The number of regions must be less than or equal to 50

  6. Programs for the automatic gamma-ray measurement with CANBERRA 8100/QUANTA system

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Sakai, Eiji; Kubo, Katsumi.

    1982-07-01

    Some programs have been prepared for the automatic operation of the CANBERRA 8100/QUANTA System for the gamma-ray spectrum measurement. The main parts of these programs are: (1) to collect and record on magnetic disks the data of gamma-ray spectra automatically, while the recorded data are analyzed to estimate the nuclides which generate photopeaks of spectra and to calculate those concentrations; (2) to draw plotted diagrams of pulse height distributions of gamma-ray spectra data and other data by the additional digital plotter; and etc. (author)

  7. Calculation of angular distribution of 662 keV gamma rays by Monte Carlo method in copper medium

    International Nuclear Information System (INIS)

    Kahraman, A.; Ozmutlu, E.N.; Gurler, O.; Yalcin, S.; Kaynak, G.; Gundogdu, O.

    2009-01-01

    This paper presents results on the angular distribution of Compton scattering of 662 keV gamma photons in both forward and backward hemispheres in copper medium. The number of scattered events graph has been determined for scattered gamma photons in both the forward and backward hemispheres and theoretical saturation thicknesses have been obtained using these results. Furthermore, response function of a 51x51 mm NaI(Tl) detector at 60 deg. angle with incoming photons scattered from a 10 mm thick copper layer has been determined using Monte Carlo method.

  8. Calculation of angular distribution of 662 keV gamma rays by Monte Carlo method in copper medium

    Energy Technology Data Exchange (ETDEWEB)

    Kahraman, A.; Ozmutlu, E.N. [Physics Department, Faculty of Arts and Sciences, Uludag University, Gorukle Campus, 16059 Bursa (Turkey); Gurler, O. [Physics Department, Faculty of Arts and Sciences, Uludag University, Gorukle Campus, 16059 Bursa (Turkey)], E-mail: ogurler@uludag.edu.tr; Yalcin, S. [Kastamonu University, Education Faculty, 37200 Kastamonu (Turkey); Kaynak, G. [Physics Department, Faculty of Arts and Sciences, Uludag University, Gorukle Campus, 16059 Bursa (Turkey); Gundogdu, O. [NCCPM, Medical Physics, Royal Surrey County, Hospital, GU2 7XX (United Kingdom); University of Kocaeli, Umuttepe Campus, 41100 Kocaeli (Turkey)

    2009-12-15

    This paper presents results on the angular distribution of Compton scattering of 662 keV gamma photons in both forward and backward hemispheres in copper medium. The number of scattered events graph has been determined for scattered gamma photons in both the forward and backward hemispheres and theoretical saturation thicknesses have been obtained using these results. Furthermore, response function of a 51x51 mm NaI(Tl) detector at 60 deg. angle with incoming photons scattered from a 10 mm thick copper layer has been determined using Monte Carlo method.

  9. Mixing of ground-state rotational and gamma and beta vibrational bands in the region A>=228

    Energy Technology Data Exchange (ETDEWEB)

    Mittal, R; Sahota, H S [Punjabi Univ., Patiala (India). Dept. of Physics

    1983-06-21

    The mixing of beta, gamma and ground-state bands has been investigated through the experimental determination of mixing parameters Zsub(..gamma..) and Zsub(..beta gamma..). These Zsub(..gamma..) values have been compared with the theoretical calculations of this parameter from the solutions of time-dependent HFB equations on the adiabatic and nonadiabatic assumptions. The experimental values are in better agreement with the results obtained under the nonadiabatic assumption, valid for small deviations from the spherical symmetry.

  10. Neutron and gamma characterization within the FFTF reactor cavity

    International Nuclear Information System (INIS)

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

  11. Requirements to a Norwegian national automatic gamma monitoring system

    DEFF Research Database (Denmark)

    Lauritzen, B.; Jensen, Per Hedemann; Nielsen, F.

    2005-01-01

    increments above the natural background levels. The study is based upon simplified deterministic calculations of the radiological consequences of generic nuclear accident scenarios. The density of gammamonitoring stations has been estimated from an analysis of the dispersion of radioactive materials over......An assessment of the overall requirements to a Norwegian gamma-monitoring network is undertaken with special emphasis on the geographical distribution of automatic gamma monitoring stations, type of detectors in such stations and the sensitivity of thesystem in terms of ambient dose equivalent rate...

  12. Precipitation kinetics of the phase. gamma. ' in Fe-Ni-Cr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.A.; Pozarnik, F.

    1984-04-01

    The authors investigated the precipitation and coalescence kinetics of the ..gamma..'phase in alloy 800, an austenitic steel with 33% Ni, 20% Cr and small amounts of Ti and Al. The results led to a law concerning the variation with temperature, ageing, and chemical composition of the particle size in the ..gamma..'phase. This law was used to calculate the variation of the elasticity limit of the alloy due to the formation of the ..gamma..'phase. The calculations were based on the theories of interaction of (weakly and strongly coupled) dislocation pairs with coherent particles ordered without constraint; the anisotropy of tension along the dislocation line was taken into account as well as the influence of the deformation induced by the misfit. A comparison with experimental results shows that averaging does not occur until 2x10/sup 5/ h at operating temperatures below 800 K.

  13. The effect of gamma dose on the PADC detectors

    International Nuclear Information System (INIS)

    Zaky, M.F.; Youssef, A.A.

    2002-01-01

    The effect of irradiation by 6 0C O gamma rays in the range 0-60 K gray has been examined on CR-39 SSNTDs. The fission fragment tracks diameter were measured using an optical microscope, the bulk etching rate was calculated using the equation V B = D/2 t. The results indicate that, the track diameter is seen increase slowly in the range 0-60 K gray. The bulk etching rate increases almost linearly as the given gamma dose increases up to (22.5 K Gray), at higher doses the bulk etching rate increases exponentially. The exposure of the CR-39 to gamma rays could sensitize the CR-39 plastic and thus improve the Z/P threshold for track registration

  14. A new method for studying the transport of gamma photons in various geological materials by combining the SSNTD technique with Monte Carlo simulations

    International Nuclear Information System (INIS)

    Misdaq, M.A.; Merzouki, A.; Bourzik, W.; Sfairi, T.

    2000-01-01

    The gamma dose rate due to the uranium and thorium series as well as the potassium 40 nuclei represents a large fraction of the total dose rate from the natural background. Natural gamma-activities of rock and soil samples collected from volcanic areas have been determined using gamma-ray spectrometry. The corresponding gamma dose rates in air have been measured by means of thermoluminescence (TL) dosimeters. Annual absorbed gamma dose rates have been evaluated in different soil samples belonging to an archaeological site by using experimental and calculational methods. Uranium and thorium contents in different geological samples have been determined by using CR-39 and LR-115 type II solid state nuclear track detectors (SSNTD) and calculating the probabilities for alpha particles emitted by the uranium and thorium series to reach and be registered on the SSNTD films. A new method has been developed based on calculating the self-absorption and transmission coefficients of the gamma photons emitted by the uranium and thorium families as well as the potassium 40 isotope for evaluating the gamma dose rate in the considered geological samples. Transport of gamma-photons across parallelepipedic blocks of the geological materials studied has been investigated. Gamma dose rates have been evaluated in the atmosphere of different geological deposits. (author)

  15. On the regularities of gamma-ray initiated emission of really-secondary electrons

    International Nuclear Information System (INIS)

    Grudskij, M.Ya.; Roldugin, N.N.; Smirnov, V.V.

    1982-01-01

    Emission regularities of the really-secondary electrons from metals are discussed on the basis of experimental data on electron emission characteristics under gamma radiation of incident quanta produced for a wide energy range (Esub(γ)=0.03+-2 MeV) and atomic numbers of target materials (Z=13+-79). Comparison with published experimental and calculated data is performed. It is shown that yield of the really-secondary electrons into vacuum from the target surface bombarded with a normally incident collimated beam of gamma radiation calculating on energy unit absorbed in the yield zone of the really-secondary electrons is determined only with the target material emittivity and can be calculated if spatial-energy distributions and the number of secondary fast electrons emitted out of the target are known

  16. Methods of bone marrow dose calculation

    International Nuclear Information System (INIS)

    Taboaco, R.C.

    1982-02-01

    Several methods of bone marrow dose calculation for photon irradiation were analised. After a critical analysis, the author proposes the adoption, by the Instituto de Radioprotecao e Dosimetria/CNEN, of Rosenstein's method for dose calculations in Radiodiagnostic examinations and Kramer's method in case of occupational irradiation. It was verified by Eckerman and Simpson that for monoenergetic gamma emitters uniformly distributed within the bone mineral of the skeleton the dose in the bone surface can be several times higher than dose in skeleton. In this way, is also proposed the Calculation of tissue-air ratios for bone surfaces in some irradiation geometries and photon energies to be included in the Rosenstein's method for organ dose calculation in Radiodiagnostic examinations. (Author) [pt

  17. Numerical simulations on efficiency and measurement of capabilities of BGO detectors for high energy gamma ray

    CERN Document Server

    Wen Wan Xin

    2002-01-01

    The energy resolution and time resolution of two phi 75 x 100 BGO detectors for high energy gamma ray newly made were measured with sup 1 sup 3 sup 7 Cs and sup 6 sup 0 Co resources. The two characteristic gamma rays of high energy emitted from the thermal neutron capture of germanium in BGO crystal were used for the energy calibration of gamma spectra. The intrinsic photopeak efficiency, single escape probability and double escape probabilities of BGO detectors in photon energy range of 4-30 MeV are numerically calculated with GEANT code. The real count response and count ratio of the uniformly distributed incident photons in energy range of 0-30 MeV are also calculated. The distortion of gamma spectra caused by the photon energy loss extension to lower energy in detection medium is discussed

  18. APPLE-2: an improved version of APPLE code for plotting neutron and gamma ray spectra and reaction rates

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi.

    1982-07-01

    A computer code APPLE-2 which plots the spatial distribution of energy spectra of multi-group neutron and/or gamma ray fluxes, and reaction rates has been developed. This code is an improved version of the previously developed APPLE code and has the following features: (1) It plots energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT and MORSE. (2) It calculates and plots the spatial distribution of neutron and gamma ray fluxes and various types of reaction rates such as nuclear heating rates, operational dose rates, displacement damage rates. (3) Input data specification is greatly simplified by the use of standard, response libraries and by close coupling with radiation transport calculation codes. (4) Plotting outputs are given in camera ready form. (author)

  19. SWEPP Gamma-Ray Spectrometer System software design description

    International Nuclear Information System (INIS)

    Femec, D.A.; Killian, E.W.

    1994-08-01

    To assist in the characterization of the radiological contents of contract-handled waste containers at the Stored Waste Examination Pilot Plant (SWEPP), the SWEPP Gamma-Ray Spectrometer (SGRS) System has been developed by the Radiation Measurements and Development Unit of the Idaho National Engineering Laboratory. The SGRS system software controls turntable and detector system activities. In addition to determining the concentrations of gamma-ray-emitting radionuclides, this software also calculates attenuation-corrected isotopic mass ratios of-specific interest. This document describes the software design for the data acquisition and analysis software associated with the SGRS system

  20. SWEPP Gamma-Ray Spectrometer System software design description

    Energy Technology Data Exchange (ETDEWEB)

    Femec, D.A.; Killian, E.W.

    1994-08-01

    To assist in the characterization of the radiological contents of contract-handled waste containers at the Stored Waste Examination Pilot Plant (SWEPP), the SWEPP Gamma-Ray Spectrometer (SGRS) System has been developed by the Radiation Measurements and Development Unit of the Idaho National Engineering Laboratory. The SGRS system software controls turntable and detector system activities. In addition to determining the concentrations of gamma-ray-emitting radionuclides, this software also calculates attenuation-corrected isotopic mass ratios of-specific interest. This document describes the software design for the data acquisition and analysis software associated with the SGRS system.

  1. Parameters calculation of shielding experiment

    International Nuclear Information System (INIS)

    Gavazza, S.

    1986-02-01

    The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

  2. J/psi-> gamma B anti B decays and the quark-pair creation model

    CERN Document Server

    Ping Rong Gang; Shen Peng Nian; Zou Bing Song

    2002-01-01

    The authors generalize the quark-pair creation model to a study of the radiative decays J/psi-> gamma B anti B by assuming that the u, d or s quark pairs are created with the same interaction strength. From the calculation of the ratio of the decay widths GAMMA(J/psi-> gamma p anti B)/GAMMA(J/psi->p anti p), the authors extract the quark-pair creation strength gI=15.40 GeV. Based on the SU(6) spin-flavour basis and the 'uds' basis, the radiative decay branching ratios containing strange baryons are evaluated. Measurements for these decay widths from the BESII data are suggested

  3. Gamma ray attenuation coefficient measurement for neutron-absorbent materials

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, Majid [Isfahan Nuclear Science and Technology Research Institute (NSTRT), Reactor and Accelerators Research and Development School, Atomic Energy Organization (Iran, Islamic Republic of)], E-mail: m_jalali@entc.org.ir; Mohammadi, Ali [Faculty of Science, Department of Physics, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2008-05-15

    The compounds Na{sub 2}B{sub 4}O{sub 7}, H{sub 3}BO{sub 3}, CdCl{sub 2} and NaCl and their solutions attenuate gamma rays in addition to neutron absorption. These compounds are widely used in the shielding of neutron sources, reactor control and neutron converters. Mass attenuation coefficients of gamma related to the four compounds aforementioned, in energies 662, 778.9, 867.38, 964.1, 1085.9, 1173, 1212.9, 1299.1,1332 and 1408 keV, have been determined by the {gamma} rays transmission method in a good geometry setup; also, these coefficients were calculated by MCNP code. A comparison between experiments, simulations and Xcom code has shown that the study has potential application for determining the attenuation coefficient of various compound materials. Experiment and computation show that H{sub 3}BO{sub 3} with the lowest average Z has the highest gamma ray attenuation coefficient among the aforementioned compounds.

  4. Examination of irradiated fuel elements using gamma scanning technique

    International Nuclear Information System (INIS)

    Ichim, O.; Mincu, M.; Man, I.; Stanica, M.

    2016-01-01

    The purpose of this paper is to validate the gamma scanning technique used to calculate the activity of gamma fission products from CANDU/TRIGA irradiated fuel elements. After a short presentation of the equipments used and their characteristics, the paper describes the calibration technique for the devices and how computed tomography reconstruction is done. Following the previously mentioned steps is possible to obtain the axial and radial profiles and the computed tomography reconstruction for calibration sources and for the irradiated fuel elements. The results are used to validate the gamma scanning techniques as a non-destructive examination method. The gamma scanning techniques will be used to: identify the fission products in the irradiated CANDU/TRIGA fuel elements, construct the axial and radial distributions of fission products, get the distribution in cross section through computed tomography reconstruction, and determine the nuclei number and the fission products activity of the irradiated CANDU/TRIGA fuel elements. (authors)

  5. Characterizing the source properties of terrestrial gamma ray flashes

    Science.gov (United States)

    Dwyer, Joseph R.; Liu, Ningyu; Eric Grove, J.; Rassoul, Hamid; Smith, David M.

    2017-08-01

    Monte Carlo simulations are used to determine source properties of terrestrial gamma ray flashes (TGFs) as a function of atmospheric column depth and beaming geometry. The total mass per unit area traversed by all the runaway electrons (i.e., the total grammage) during a TGF, Ξ, is introduced, defined to be the total distance traveled by all the runaway electrons along the electric field lines multiplied by the local air mass density along their paths. It is shown that key properties of TGFs may be directly calculated from Ξ and its time derivative, including the gamma ray emission rate, the current moment, and the optical power of the TGF. For the calculations presented in this paper, a standard TGF gamma ray fluence, F0 = 0.1 cm-2 above 100 keV for a spacecraft altitude of 500 km, and a standard total grammage, Ξ0 = 1018 g/cm2, are introduced, and results are presented in terms of these values. In particular, the current moments caused by the runaway electrons and their accompanying ionization are found for a standard TGF fluence, as a function of source altitude and beaming geometry, allowing a direct comparison between the gamma rays measured in low-Earth orbit and the VLF-LF radio frequency emissions recorded on the ground. Such comparisons should help test and constrain TGF models and help identify the roles of lightning leaders and streamers in the production of TGFs.

  6. Radiation Build-Up Of High Energy Gamma In Shielding Of High Atomic Number

    International Nuclear Information System (INIS)

    Yuliati, Helfi; Akhadi, Mukhlis

    2000-01-01

    Research to observe effect of radiation build-up factor (b) in iron (Fe) and lead (Pb) for high energy gamma shielding from exp.137 Cs (E gamma : 662 keV) and exp.60 Co (E gamma : 1332 keV) sources has been carried out. Research was conducted bt counting of radiation intensity behind shielding with its thickness vary from 1 to 5 times of half value thickness (HVT). NaI (TI) detector which connected to multi channel analyzer (MCA) was used for the counting. Calculation result show that all of b value are near to 1 (b∼1) both for Fe and Pb. Without inserting b in calculation, from the experiment it was obtained HVT value of Fe for high gamma radiation of 662 and 1332 keV were : (12,94 n 0,03) mm and (17,33 n 0,01) mm with their deviation standards were 0,2% and 0,06% respectively. Value of HVT for Pb with the same energy were : (6,31 n 0,03) mm and (11,86 n 0,03) mm with their deviation standars were : 0,48% and 0,25% respectively. HVL concept could be applied directly to estimate shielding thickness of high atomic number of high energy gamma radiation, without inserting correction of radiation build-up factor

  7. Gamma ray beam transmutation

    International Nuclear Information System (INIS)

    Imasaki, K.; Li, D.; Miyamoto, S.; Amano, S.; Motizuki, T.

    2007-01-01

    We have proposed a new approach to nuclear transmutation by a gamma ray beam of Compton scattered laser photon. We obtained 20 MeV gamma ray in this way to obtain transmutation rates with the giant resonance of 1 97Au and 1 29Iodine. The rate of the transmutation agreed with the theoretical calculation. Experiments on energy spectrum of positron, electron and neutron from targets were performed for the energy balance and design of the system scheme. The reaction rate was about 1.5∼4% for appropriate photon energies and neutron production rate was up to 4% in the measurements. We had stored laser photon more than 5000 times in a small cavity which implied for a significant improvement of system efficiency. Using these technologies, we have designed an actual transmutation system for 1 29Iodine which has a 16 million year's activity. In my presentation, I will address the properties of this scheme, experiments results and transmutation system for iodine transmutation

  8. Application of gamma-spectrometry to post irradiation examination

    International Nuclear Information System (INIS)

    Kim, S.K.; Huh, Y.H.; Park, K.J.

    1982-01-01

    A areat variety of nuclear gamma rays emitted from fission and activation products of spent nuclear fuel contains much information that can be elicited without affecting the integrity of the fuel elements. In the present work, a versatile code CAERI was developed which locates peaks and calculates their areas for X-rays as well as gamma rays using elegant features of some widely used programs for gamma ray peak fitting. CAERI coded in FORTRAN used infinite series approximation more accurate than other workers' various, simple, piecewise series approximations for evaluation of the Voigt function which represents the X-ray peak with nonneglible natural line width. CAERI can handle even a complex multiplet consisting of peaks from X-rays and rays in arbitrary mixture, which one often encounters in the isotopic analysis of heavy elements such as U and Pu. (Author)

  9. Beta and gamma decay heat evaluation for the thermal fission of 235U

    International Nuclear Information System (INIS)

    Schenter, G.K.; Schmittroth, F.

    1979-01-01

    Beta and gamma fission product decay heat curves are evaluated for the thermal fission of 235 U. Experimental data that include beta, gamma, and total measurements are combined with summation calculations based on ENDF/B in a consistent evaluation. Least-squares methods are used that take proper account of data uncertainties and correlations. 4 figures, 2 tables

  10. Determination of radionuclides in environmental test items at CPHR: traceability and uncertainty calculation.

    Science.gov (United States)

    Carrazana González, J; Fernández, I M; Capote Ferrera, E; Rodríguez Castro, G

    2008-11-01

    Information about how the laboratory of Centro de Protección e Higiene de las Radiaciones (CPHR), Cuba establishes its traceability to the International System of Units for the measurement of radionuclides in environmental test items is presented. A comparison among different methodologies of uncertainty calculation, including an analysis of the feasibility of using the Kragten-spreadsheet approach, is shown. In the specific case of the gamma spectrometric assay, the influence of each parameter, and the identification of the major contributor, in the relative difference between the methods of uncertainty calculation (Kragten and partial derivative) is described. The reliability of the uncertainty calculation results reported by the commercial software Gamma 2000 from Silena is analyzed.

  11. Determination of radionuclides in environmental test items at CPHR: Traceability and uncertainty calculation

    International Nuclear Information System (INIS)

    Carrazana Gonzalez, J.; Fernandez, I.M.; Capote Ferrera, E.; Rodriguez Castro, G.

    2008-01-01

    Information about how the laboratory of Centro de Proteccion e Higiene de las Radiaciones (CPHR), Cuba establishes its traceability to the International System of Units for the measurement of radionuclides in environmental test items is presented. A comparison among different methodologies of uncertainty calculation, including an analysis of the feasibility of using the Kragten-spreadsheet approach, is shown. In the specific case of the gamma spectrometric assay, the influence of each parameter, and the identification of the major contributor, in the relative difference between the methods of uncertainty calculation (Kragten and partial derivative) is described. The reliability of the uncertainty calculation results reported by the commercial software Gamma 2000 from Silena is analyzed

  12. A new computationally-efficient computer program for simulating spectral gamma-ray logs

    International Nuclear Information System (INIS)

    Conaway, J.G.

    1995-01-01

    Several techniques to improve the accuracy of radionuclide concentration estimates as a function of depth from gamma-ray logs have appeared in the literature. Much of that work was driven by interest in uranium as an economic mineral. More recently, the problem of mapping and monitoring artificial gamma-emitting contaminants in the ground has rekindled interest in improving the accuracy of radioelement concentration estimates from gamma-ray logs. We are looking at new approaches to accomplishing such improvements. The first step in this effort has been to develop a new computational model of a spectral gamma-ray logging sonde in a borehole environment. The model supports attenuation in any combination of materials arranged in 2-D cylindrical geometry, including any combination of attenuating materials in the borehole, formation, and logging sonde. The model can also handle any distribution of sources in the formation. The model considers unscattered radiation only, as represented by the background-corrected area under a given spectral photopeak as a function of depth. Benchmark calculations using the standard Monte Carlo model MCNP show excellent agreement with total gamma flux estimates with a computation time of about 0.01% of the time required for the MCNP calculations. This model lacks the flexibility of MCNP, although for this application a great deal can be accomplished without that flexibility

  13. Whole-procedure clinical accuracy of Gamma Knife treatments of large lesions

    International Nuclear Information System (INIS)

    Ma Lijun; Chuang, Cynthia; Descovich, Martina; Petti, Paula; Smith, Vernon; Verhey, Lynn

    2008-01-01

    The mechanical accuracy of Gamma Knife radiosurgery based on single-isocenter measurement has been established to within 0.3 mm. However, the full delivery accuracy for Gamma Knife treatments of large lesions has only been estimated via the quadrature-sum analysis. In this study, the authors directly measured the whole-procedure accuracy for Gamma Knife treatments of large lesions to examine the validity of such estimation. The measurements were conducted on a head-phantom simulating the whole treatment procedure that included frame placement, computed tomography imaging, treatment planning, and treatment delivery. The results of the measurements were compared with the dose calculations from the treatment planning system. Average agreements of 0.1-1.6 mm for the isodose lines ranging from 25% to 90% of the maximum dose were found despite potentially large contributing uncertainties such as 3-mm imaging resolution, 2-mm dose grid size, 1-mm frame registration, multi-isocenter deliveries, etc. The results of our measurements were found to be significantly smaller (>50%) than the calculated value based on the quadrature-sum analysis. In conclusion, Gamma Knife treatments of large lesions can be delivered much more accurately than predicted from the quadrature-sum analysis of major sources of uncertainties from each step of the delivery chain.

  14. Comptonization of gamma rays by cold electrons

    International Nuclear Information System (INIS)

    Xu, Yueming; Ross, R.R.; Mccray, R.

    1991-01-01

    An analytic method is developed for calculating the emergent spectrum of gamma-rays and X-rays scattered in a homogeneous medium with low-temperature electrons. The Klein-Nishina corrections of the scattering cross section and absorption processes are taken in account. The wavelength relaxation and the spatial diffusion problems are solved separately, and the emergent spectrum is calculated by convolving the evolution function of the spectrum in an infinite medium with the photon luminosity resulting from the spatial diffusion in a finite sphere. The analytic results are compared with that of Monte Carlo calculations and it is concluded that the analytic result is quite accurate. 9 refs

  15. Dose Rate and Mass Attenuation Coefficients of Gamma Ray for Concretes

    CERN Document Server

    Abdel-Latif, A A; Kansouh, W A; El-Sayed, F H

    2003-01-01

    This work is concerned with the study of the leakage gamma ray dose and mass attenuation coefficients for ordinary, basalt and dolomite concretes made from local ores. Concretes under investigation were constructed from gravel, basalt and dolomite ores, and then reconstructed with the addition of 3% steel fibers by weight. Measurements were carried out using a collimated beam from sup 6 sup 0 Co gamma ray source and sodium iodide (3x3) crystal with the genie 2000 gamma spectrometer. The obtained fluxes were transformed to gamma ray doses and displayed in the form of gamma ray dose rates distribution. The displayed curves were used to estimate the linear attenuation coefficients (mu), the relaxation lengths (lambda), half value layer (t sub 1 /2) and tenth value layer (t sub 1 /10). Also, The total mass attenuation coefficients of gamma ray have been calculated to the concerned concretes using XCOM (version 3.1) program and database elements cross sections from Z=1 to 100 at energies from 10 keV to 100 MeV. In...

  16. Initial characterization of the ATR [Advanced Test Reactor] Large Gamma Facility

    International Nuclear Information System (INIS)

    Schnitzler, B.G.; Rogers, J.W.

    1986-05-01

    Radiation fields in the ATR Large Gamma Facility test volume are characterized. The preliminary characterization efforts described in this report include total dose rate measurements in the facility, development of a simple methodology for calculating radiation fields from the ATR fuel element power histories, and a comparison of the measured and calculated values

  17. Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B.W.; Firestone, Richard B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H.D.

    2010-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF has been used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy an is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We use CASINO, a version of DICEBOX that is modified for this purpose. This can be used to simulate the neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modelling of unknown assemblies.

  18. Saturation and porosity measurements of different soil samples by gamma ray transmission

    International Nuclear Information System (INIS)

    Akbal, S.; Filiz Baytas, A.

    2000-01-01

    Gamma-ray transmission methods have been used accurately for the study of the properties of soil samples. In this study, the soil samples were collected from various regions of Turkey and a Nal (TI) detector measured the attenuation of strongly collimated monoenergetic gamma beam (from Cs-137) through soil samples. The water saturation and porosity were therefore calculated from the transmission measurements for each soil sample. (authors)

  19. Two-dimensional sensitivity calculation code: SENSETWO

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.

    1979-05-01

    A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)

  20. Monte Carlo calculations of neutron and gamm-ray energy spectra for fusion-reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.

    1983-08-01

    Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE