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
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
Rivard, M J; D'Errico, F; Tsai, J S; Ulin, K; Engler, M J
2002-01-01
The sup 2 sup 5 sup 2 Cf neutron air kerma strength conversion factor (S sub K sub N /m sub C sub f) is a parameter needed to convert the radionuclide mass (mu g) provided by Oak Ridge National Laboratory into neutron air kerma strength required by modern clinical brachytherapy dosimetry formalisms indicated by Task Group No. 43 of the American Association of Physicists in Medicine (AAPM). The impact of currently used or proposed encapsulating materials for sup 2 sup 5 sup 2 Cf brachytherapy sources (Pt/Ir-10%, 316L stainless steel, nitinol, and Zircaloy-2) on S sub K sub N /m sub C sub f was calculated and results were fit to linear equations. Only for substantial encapsulation thicknesses, did S sub K sub N /m sub C sub f decrease, while the impact of source encapsulation composition is increasingly negligible as Z increases. These findings are explained on the basis of the non-relativistic kinematics governing the majority of sup 2 sup 5 sup 2 Cf neutron interactions. Neutron kerma and energy spectra resul...
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)
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
Fission neutron multiplicity calculations
International Nuclear Information System (INIS)
Maerten, H.; Ruben, A.; Seeliger, D.
1991-01-01
A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs
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
Calculations of neutron spectra after neutron-neutron scattering
Energy Technology Data Exchange (ETDEWEB)
Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Howell, C R [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G E [North Carolina State University, Raleigh, NC 27695-8202 (United States); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V N [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation)
2004-09-01
A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.
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)
Neutron quality parameters versus energy below 4 MeV from microdosimetric calculations
International Nuclear Information System (INIS)
Stinchcomb, T.G.; Borak, T.B.
1983-01-01
Charged-particle production by neutrons and the resulting energy-deposition spectra in micron-sized spheres of tissue of varying diameters were calculated from thermal energies to 4 MeV. These data were used to obtain dose-average values of several quality-indicating parameters as functions of neutron energy and of tissue sphere diameter. The contrast among the parameters is shown and discussed. Applications are made to two neutron spectra, one a fission spectrum in air and the other a moderated spectrum at the center of an irradiated cube of water
International Nuclear Information System (INIS)
Sekimoto, H.
1987-01-01
The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source
Calculation of neutron kerma in tissues
International Nuclear Information System (INIS)
Vega C, H.R.; Manzanares A, E.
2004-01-01
Neutron kerma of normal and tumor tissues has been calculated using the tissues elemental concentration. A program developed in Math cad contains the kerma factors of C, H, O, N, Na, Mg, P, S, Cl, K, etc. that are in normal and tumor human tissues. Having the elemental composition of any human tissue the neutron kerma can be calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic, also neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large. (Author)
Neutronics calculation of RTP core
Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.
2017-01-01
Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.
Neutron and gamma-ray transport experiments in liquid air
International Nuclear Information System (INIS)
Farley, W.E.
1976-01-01
Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
Villarino, Eduardo; Hergenreder, Daniel
2000-01-01
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
Calculations on neutron irradiation damage in reactor materials
International Nuclear Information System (INIS)
Sone, Kazuho; Shiraishi, Kensuke
1976-01-01
Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)
Some neutronic calculations for KENS-II
International Nuclear Information System (INIS)
Kiyanagi, Y.; Arai, M.; Watanabe, N.
1989-01-01
Proton energies of the intense spallation neutron sources currently in operation or designed are in the range Ep ≤ 1.1 GeV. Optimization studies of the target station have so far been performed for these proton energies. The KENS-II project has been included in the Japanese Hadron Facility Project where the proton accelerator, a so-called First Ring is shared with Meson Arena for nuclear physics and μSR experiments. The possible highest proton energy for this accelerator is 2 GeV, which is the highest among the world's spallation neutron sources. The authors, therefore, performed some neutronic calculations with 2 GeV protons in order to have a good knowledge of the neutronic characteristics and the optimal parameters of the target station for KENS-II. The fraction of slow neutron intensity versus the proton energy becomes 0.8 for 2 GeV compared to that for 0.8 GeV, and this is higher than 0.67 calculated for source neutrons. The uranium target has a higher neutron productivity, 1.5 times that of the tungsten target, even for 2 GeV protons. The target radius and the moderator axial position have definite optimal values for 2 GeV protons in spite of the broader distribution of the source neutrons in target, and these are essentially similar to the results for 0.8 GeV protons. The broad distribution with a little increase in the maximum luminosity of source neutrons for 2 GeV protons could make it easier to remove the heat load from the target than the case for the same beam-power with lower energy and higher proton current. Therefore, they could conclude that the 2 GeV protons for KENS-II do not have significant difficulties in producing slow neutrons, and that non-fissile material has higher advantages to produce neutrons for higher proton energies. Detailed neutronic calculations are now under way to design a neutron target station for KENS-II. 5 refs., 10 figs
Developing neutronics calculation tools for MYRRHA
International Nuclear Information System (INIS)
Van den Eynde, G.
2006-01-01
The design of the Accelerator Driven System MYRRHA requires adequate and specialised tools in the field of neutronics calculations. In order to fill the gaps, several PhD programmes were launched. In 2005 three such PhD projects were running. Each of them focuses on different stages in the computation of a core of MYRRHA. The first project I mprovements of the spallation reaction model , a collaboration with the University of Liege, deals with the characterisation of the spallation neutron source using the INCL (Intra-Nuclear Cascade of Liege) model. Since at high energies, nuclear data are sparse, calculations rely on models. Especially for spallation reactions that occur at proton energies of several hundreds of MeV, models are the only means to evaluate the spallation source in MYRRHA. The second project 'Neutron transport with anisotropic scattering', a collaboration with the Universite Libre de Bruxelles, works on the development of a neutronics code, CASE-BSM, for systems with highly anisotropic scattering. The presence in large amounts of both lead and bismuth atoms in the MYRRHA core results in a highly anisotropic scattering of the neutrons in the bulk of the coolant. Neglecting this effect has large consequences on both global parameters, like keff, as well as on local parameters, like the neutron flux seen by the vessel. The third project, 'ALEPH: An integrated Monte Carlo bun-up tool', a collaboration with Ghent University, treats the last phase of a core calculation: the depletion of the fuel during irradiation. For an experimental machine like MYRRHA it is of utmost importance to have a fast calculational tool to evaluate the incineration of both isotopes present in the fuel as isotopes present in experimental devices. The main objective is to improve the current quality of the neutronics codes focused on ADS applications and to have this knowledge 'in-house'
FENDL neutronics benchmark: Specifications for the calculational neutronics and shielding benchmark
International Nuclear Information System (INIS)
Sawan, M.E.
1994-12-01
During the IAEA Advisory Group Meeting on ''Improved Evaluations and Integral Data Testing for FENDL'' held in Garching near Munich, Germany in the period 12-16 September 1994, the Working Group II on ''Experimental and Calculational Benchmarks on Fusion Neutronics for ITER'' recommended that a calculational benchmark representative of the ITER design should be developed. This report describes the neutronics and shielding calculational benchmark available for scientists interested in performing analysis for this benchmark. (author)
Relativistic mean field calculations in neutron-rich nuclei
Energy Technology Data Exchange (ETDEWEB)
Gangopadhyay, G.; Bhattacharya, Madhubrata [Department of Physics, University of Calcutta, 92 Acharya Prafulla Chandra Road, Kolkata 700 009 (India); Roy, Subinit [Saha Institute of Nuclear Physics, Block AF, Sector 1, Kolkata- 700 064 (India)
2014-08-14
Relativistic mean field calculations have been employed to study neutron rich nuclei. The Lagrange's equations have been solved in the co-ordinate space. The effect of the continuum has been effectively taken into account through the method of resonant continuum. It is found that BCS approximation performs as well as a more involved Relativistic Continuum Hartree Bogoliubov approach. Calculations reveal the possibility of modification of magic numbers in neutron rich nuclei. Calculation for low energy proton scattering cross sections shows that the present approach reproduces the density in very light neutron rich nuclei.
Uncertainty analysis of neutron transport calculation
International Nuclear Information System (INIS)
Oka, Y.; Furuta, K.; Kondo, S.
1987-01-01
A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)
Calculations of accelerator-based neutron sources characteristics
International Nuclear Information System (INIS)
Tertytchnyi, R.G.; Shorin, V.S.
2000-01-01
Accelerator-based quasi-monoenergetic neutron sources (T(p,n), D(d;n), T(d;n) and Li (p,n)-reactions) are widely used in experiments on measuring the interaction cross-sections of fast neutrons with nuclei. The present work represents the code for calculation of the yields and spectra of neutrons generated in (p, n)- and ( d; n)-reactions on some targets of light nuclei (D, T; 7 Li). The peculiarities of the stopping processes of charged particles (with incident energy up to 15 MeV) in multilayer and multicomponent targets are taken into account. The code version is made in terms of the 'SOURCE,' a subroutine for the well-known MCNP code. Some calculation results for the most popular accelerator- based neutron sources are given. (authors)
Sensitivity of neutron air transport to nitrogen cross section uncertainties
International Nuclear Information System (INIS)
Niiler, A.; Beverly, W.B.; Banks, N.E.
1975-01-01
The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures
Possibilities of delayed neutron fraction (βeff) calculation and measurement
International Nuclear Information System (INIS)
Michalek, S.; Hascik, J.; Farkas, G.
2008-01-01
The influence of the delayed neutrons on the reactor dynamics can be understood through their impact on the reactor power change rate. In spite of the fact that delayed neutrons constitute only a very small fraction of the total number of neutrons generated from fission, they play a dominant role in the fission chain reaction control. If only the prompt neutrons existed, the reactor operation would become impossible due to the fast reactor power changes. The exact determination of delayed neutrons main parameter, the delayed neutron fraction (β eff ), is very important in the field of reactor physics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of calculations and experiments. In consequence of difficulties in β eff experimental measurement, this value in exact state use to be determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. Determination of β eff requires criticality calculations. In the past, k eff used to be traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum- weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. In this work, a summary of possible β eff calculation methods can be found and a calculation of β eff for VR-1 training reactor in one operation state is made using the prompt method, by MCNP5 code. Also a method of β eff kinetic measurement on VR-1 training reactor at Czech Technical University in Prague using in-pile kinetic technique is outlined (authors)
Average cross sections calculated in various neutron fields
International Nuclear Information System (INIS)
Shibata, Keiichi
2002-01-01
Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)
Early fusion reactor neutronic calculations: A reevaluation
International Nuclear Information System (INIS)
Perry, R.T.
1996-01-01
Several fusion power plant design studies were made at a number of universities and laboratories in the late 1960s and early 1970s. These studies included such designs as the Princeton Plasma Physics Laboratory Fusion Power Plan and the University of Wisconsin UWMAK-I Reactor Neutronic analyses of the blankets and shields were part of the studies. During this time there were dissertations written on neutronic analysis systems and the results of neutronic analysis on several blanket and shield designs. The results were presented in the literature. Now in the fifth decade of fusion research, investigators often return to the earlier analyses for the neutronic results that are applicable to current blanket and shield designs, with the idea of using the older work as a basis for the new. However, the analyses of the past were made with cross-section data sets that have long been replaced with more modern versions. In addition, approximations were often made to the cross sections used because more exact data were not available. Because these results are used as guides, it is important to know if they are reproducible using more modern data. In this paper, several of the neutronic calculations made in the early studies are repeated using the MATXS-11 data library. This library is the ENDF/B-VI version of the MATXS-5 library. The library has 80 neutron groups. Tritium breeding ratios, heating rates, and fluxes are calculated and compared. This transport code used here is the one- dimensional S n code, ONEDANT. It is important to note that the calculations here are not to be considered as benchmarks because parameter and sensitivity studies were not made. They are used only to see if the results of older calculations are in reasonable agreement with a more modern library
International Nuclear Information System (INIS)
Talamo, Alberto; Gohar, Yousry
2016-01-01
This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division
2016-06-01
This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.
Statistical theory for calculating energy spectra of β-delayed neutrons
International Nuclear Information System (INIS)
Kawano, Toshihiko; Moeller, Peter; Wilson, William B.
2008-01-01
Theoretical β-delayed neutron spectra are calculated based on the Quasi-particle Random Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after β-decay to the granddaughter residual are more accurately calculated than previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra reasonably agree with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors. (authors)
Direct Discrete Method for Neutronic Calculations
International Nuclear Information System (INIS)
Vosoughi, Naser; Akbar Salehi, Ali; Shahriari, Majid
2002-01-01
The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for a cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
Villarino, E.A.
1990-01-01
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es
Calculation of neutron background for underground experiments
Energy Technology Data Exchange (ETDEWEB)
Tomasello, V. [Department of Physics and Astronomy, University of Sheffield, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Physikalisches Institut, Eberhard Karls Universitaet Tuebingen, Auf der Morgenstelle 14, Tuebingen D-72076 (Germany)], E-mail: v.tomasello@sheffield.ac.uk; Kudryavtsev, V.A.; Robinson, M. [Department of Physics and Astronomy, University of Sheffield, Hounsfield Road, Sheffield S3 7RH (United Kingdom)
2008-10-01
New generation dark matter experiments aim at exploring the 10{sup -9}-10{sup -10}pb cross-section region for the WIMP-nucleon scalar interactions. Neutrons produced in the detector components are one of the main factors that can limit detector sensitivity. Estimation of the background from this source then becomes a crucial task for designing future large-scale detectors. Energy spectra and production rates for neutrons coming from radioactive contamination are required for all materials in and around the detector. In order to estimate neutron yields and spectra, the cross-sections of ({alpha},n) reactions and probabilities of transitions to different excited states should be known. Cross-sections and transition probabilities have been calculated using EMPIRE2.19 for several isotopes, and for some isotopes, a comparison with the experimental data is shown. The results have been used to calculate the neutron spectra from materials using the code SOURCES4A. Neutron background event rates from some detector components in a hypothetical dark matter detector based on Ge crystals have been estimated. Some requirements for the radiopurity of the materials have been deduced from the results of these simulations.
Calculation of neutron background for underground experiments
International Nuclear Information System (INIS)
Tomasello, V.; Kudryavtsev, V.A.; Robinson, M.
2008-01-01
New generation dark matter experiments aim at exploring the 10 -9 -10 -10 pb cross-section region for the WIMP-nucleon scalar interactions. Neutrons produced in the detector components are one of the main factors that can limit detector sensitivity. Estimation of the background from this source then becomes a crucial task for designing future large-scale detectors. Energy spectra and production rates for neutrons coming from radioactive contamination are required for all materials in and around the detector. In order to estimate neutron yields and spectra, the cross-sections of (α,n) reactions and probabilities of transitions to different excited states should be known. Cross-sections and transition probabilities have been calculated using EMPIRE2.19 for several isotopes, and for some isotopes, a comparison with the experimental data is shown. The results have been used to calculate the neutron spectra from materials using the code SOURCES4A. Neutron background event rates from some detector components in a hypothetical dark matter detector based on Ge crystals have been estimated. Some requirements for the radiopurity of the materials have been deduced from the results of these simulations
Calculated intensity of high-energy neutron beams
International Nuclear Information System (INIS)
Mustapha, B.; Nolen, J.A.; Back, B.B.
2004-01-01
The flux, energy and angular distributions of high-energy neutrons produced by in-flight spallation and fission of a 400 MeV/A 238 U beam and by the break-up of a 400 MeV/A deuteron beam are calculated. In both cases very intense secondary neutron beams are produced, peaking at zero degrees, with a relatively narrow energy spread. Such secondary neutron beams can be produced with the primary beams from the proposed rare isotope accelerator driver linac. The break-up of a 400 kW deuteron beam on a liquid-lithium target can produce a neutron flux of >10 10 neutrons/cm 2 /s at a distance of 10 m from the target
MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities
International Nuclear Information System (INIS)
Merchant, A.C.
1986-03-01
A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt
Measurements and calculations of neutron spectra and neutron dose distribution in human phantoms
International Nuclear Information System (INIS)
Palfalvi, J.
1984-11-01
The measurement and calculation of the radiation field around and in a phantom, with regard to the neutron component and the contaminating gamma radiation, are essential for radiation protection and radiotherapy purposes. The final report includes the development of the simple detector system, automized detector measuring facilities and a computerized evaluating system. The results of the depth dose and neutron spectra experiments and calculations in a human phantom are given
Research on neutron noise analysis stochastic simulation method for α calculation
International Nuclear Information System (INIS)
Zhong Bin; Shen Huayun; She Ruogu; Zhu Shengdong; Xiao Gang
2014-01-01
The prompt decay constant α has significant application on the physical design and safety analysis in nuclear facilities. To overcome the difficulty of a value calculation with Monte-Carlo method, and improve the precision, a new method based on the neutron noise analysis technology was presented. This method employs the stochastic simulation and the theory of neutron noise analysis technology. Firstly, the evolution of stochastic neutron was simulated by discrete-events Monte-Carlo method based on the theory of generalized Semi-Markov process, then the neutron noise in detectors was solved from neutron signal. Secondly, the neutron noise analysis methods such as Rossia method, Feynman-α method, zero-probability method, and cross-correlation method were used to calculate a value. All of the parameters used in neutron noise analysis method were calculated based on auto-adaptive arithmetic. The a value from these methods accords with each other, the largest relative deviation is 7.9%, which proves the feasibility of a calculation method based on neutron noise analysis stochastic simulation. (authors)
Subcriticality calculation in nuclear reactors with external neutron sources
Energy Technology Data Exchange (ETDEWEB)
Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br
2007-07-01
The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)
Subcriticality calculation in nuclear reactors with external neutron sources
International Nuclear Information System (INIS)
Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da
2007-01-01
The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)
International Nuclear Information System (INIS)
Bourva, L.C.A.; Croft, S.
1999-01-01
The general purpose neutron-photon-electron Monte Carlo N-Particle code, MCNP TM , has been used to simulate the neutronic characteristics of the on-site laboratory passive neutron coincidence counter to be installed, under Euratom Safeguards Directorate supervision, at the Sellafield reprocessing plant in Cumbria, UK. This detector is part of a series of nondestructive assay instruments to be installed for the accurate determination of the plutonium content of nuclear materials. The present work focuses on one aspect of this task, namely, the accurate calculation of the coincidence gate utilisation factor. This parameter is an important term in the interpretative model used to analyse the passive neutron coincidence count data acquired using pulse train deconvolution electronics based on the shift register technique. It accounts for the limited proportion of neutrons detected within the time interval for which the electronics gate is open. The Monte Carlo code MCF, presented in this work, represents a new evaluation technique for the estimation of gate utilisation factors. It uses the die-away profile of a neutron coincidence chamber generated either by MCNP TM , or by other means, to simulate the neutron detection arrival time pattern originating from independent spontaneous fission events. A shift register simulation algorithm, embedded in the MCF code, then calculates the coincidence counts scored within the electronics gate. The gate utilisation factor is then deduced by dividing the coincidence counts obtained with that obtained in the same Monte Carlo run, but for an ideal detection system with a coincidence gate utilisation factor equal to unity. The MCF code has been benchmarked against analytical results calculated for both single and double exponential die-away profiles. These results are presented along with the development of the closed form algebraic expressions for the two cases. Results of this validity check showed very good agreement. On this
Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results
International Nuclear Information System (INIS)
Madland, D.G.; LaBauve, R.J.; Nix, J.R.
1989-01-01
We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab
RAMA Methodology for the Calculation of Neutron Fluence
International Nuclear Information System (INIS)
Villescas, G.; Corchon, F.
2013-01-01
he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro
2007-02-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
Calculation of the apparent neutron parameters in a borehole geometry for neutron porosity tools
International Nuclear Information System (INIS)
Woznicka, U.; Drabina, A.
2001-01-01
This paper presents the next step of a development of the theoretical solutions, which gives a possibility to calculate the apparent neutron slowing down and migration lengths in the three region cylindrical system which represents the borehole, the intermediate zone (e.g. mud cake at the borehole walls), and the geological formation. A solution of the neutron diffusion equation in energy two-group approach for spatial moments of the neutron flux is given for the three-region cylindrical coaxial geometry. The influence of the intermediate zone is presented. The numerical code MOM3 has been written to calculate the apparent slowing down and migration lengths for the three-region cylindrical system as mentioned above. Additionally the MCNP calculation for the three-region borehole geometry is presented in the paper
Calculation of neutron importance function in fissionable assemblies using Monte Carlo method
International Nuclear Information System (INIS)
Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.
2007-01-01
The purpose of the present work is to develop an efficient solution method for the calculation of neutron importance function in fissionable assemblies for all criticality conditions, based on Monte Carlo calculations. The neutron importance function has an important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating the adjoint flux while solving the adjoint weighted transport equation based on deterministic methods. However, in complex geometries these calculations are very complicated. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on the physical concept of neutron importance has been introduced for calculating the neutron importance function in sub-critical, critical and super-critical conditions. For this propose a computer program has been developed. The results of the method have been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries. The correctness of these results has been confirmed for all three criticality conditions. Finally, the efficiency of the method for complex geometries has been shown by the calculation of neutron importance in Miniature Neutron Source Reactor (MNSR) research reactor
International Nuclear Information System (INIS)
Ruehm, Werner; Huber, Thomas; Nolte, Eckehart; Kato, Kazuo; Imanaka, Tetsuji; Egbert, Stephen D.
2005-01-01
Trace elements such as Li, B, Sm, and Gd can, despite their low elemental concentration in mineral materials, influence thermal neutron activation in Hiroshima and Nagasaki samples, due to their high thermal neutron absorption cross sections. This was demonstrated for a granite core, where the addition of those trace elements to the elemental composition of granite reduces the production of 152 Eu by some 25% at a depth of 25 cm from the surface. If typical concentrations of those trace elements are added to DS02 reference soil, however, the production of 152 Eu one meter above ground is not changed significantly, because of the high water content of the soil. This indicates that DS02 soil represents a reasonable reference material for the air-over-ground transport calculations. It must be kept in mind, however, that the local environment of any sample investigated for thermal neutron activation might be characterized by other elemental compositions. In particular, trace element and hydrogen concentrations could be considerably different from those used for DS02 reference soil. As an example it was demonstrated that in a granite gravestone thermal neutron activation of 36 Cl close to the surface might be, in the worst case, reduced by some 30%, due to increased local granite concentration in this type of environment. Beside other parameters such as, for example, individual sample geometry, the variability of trace elements in soil might be one reason for the variability that is observed in the individual thermal neutron activation measurements (Gold 1995). It is necessary, therefore, to carefully model the exposure geometry of the exposed material, its chemical composition, and the surrounding interface materials in order to obtain the best possible agreement in comparisons between calculated and measured data for thermal neutrons. (author)
Filtered thermal neutron captured cross sections measurements and decay heat calculations
International Nuclear Information System (INIS)
Pham Ngoc Son; Vuong Huu Tan
2015-01-01
Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)
Transport calculation of medium-energy protons and neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.
1978-09-01
A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)
Calculating the mass distribution of heavy nucleus fission product by neutrons
International Nuclear Information System (INIS)
Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.
1981-01-01
The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru
Calculation of 14 MeV neutron transmission
International Nuclear Information System (INIS)
Vyrskij, M.Yu.; Dubinin, A.A.; Zhuravlev, V.I.; Isaev, N.V.; Klintsov, A.A.; Krivtsov, A.S.; Linge, I.I.; Panfilov, E.I.; Prit'mov, A.P.
1979-01-01
The possibility of using the 28 group constant system (28-GCS) for calculating the transport of neutrons with initial energy of 14 MeV in thermonuclear reactor blankets is studied. A blanket project suggested by the Oak Ridge National Laboratory is used as a test version to estimate applicability of the 28-GCS. Niobium is used in a blanket as a structural material. A mixture of lithium nuclides is used for tritium production. The results of blanket test calculation and the calculational results obtained using the 28-GCS from the UKNDL library are compared. The numerical 28-group calculation of blonket is carried out by means of the ROZ-6 and ROZ-9 codes but not by the Monte-Carlo method as compared with the test calculation. Time of the blanket calculation on the BESM-6 computer by means of the ROZ-9 code in 2P 5 approximation using the 28-GCS amounts to 10 min. It is noted that to create effective codes for the numerical blanket calculation different calculational grids are necessary for different energy grups. The calculations carried out have shown the possibility of using the 28-group library of cross sections for the numerical solution of the neutron transport equation in estimating analysis of blankets
International Nuclear Information System (INIS)
Medeiros, Marcos P.C.; Rebello, Wilson F.; Andrade, Edson R.; Silva, Ademir X.
2015-01-01
Nuclear explosions are usually described in terms of its total yield and associated shock wave, thermal radiation and nuclear radiation effects. The nuclear radiation produced in such events has several components, consisting mainly of alpha and beta particles, neutrinos, X-rays, neutrons and gamma rays. For practical purposes, the radiation from a nuclear explosion is divided into i nitial nuclear radiation , referring to what is issued within one minute after the detonation, and 'residual nuclear radiation' covering everything else. The initial nuclear radiation can also be split between 'instantaneous or 'prompt' radiation, which involves neutrons and gamma rays from fission and from interactions between neutrons and nuclei of surrounding materials, and 'delayed' radiation, comprising emissions from the decay of fission products and from interactions of neutrons with nuclei of the air. This work aims at presenting isodose curves calculations at ground level by Monte Carlo simulation, allowing risk assessment and consequences modeling in radiation protection context. The isodose curves are related to neutrons produced by the prompt nuclear radiation from a hypothetical nuclear explosion with a total yield of 20 KT. Neutron fluency and emission spectrum were based on data available in the literature. Doses were calculated in the form of ambient dose equivalent due to neutrons H*(10) n - . (author)
Energy Technology Data Exchange (ETDEWEB)
Medeiros, Marcos P.C.; Rebello, Wilson F.; Andrade, Edson R., E-mail: rebello@ime.eb.br, E-mail: daltongirao@yahoo.com.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Silva, Ademir X., E-mail: ademir@nuclear.ufrj.br [Corrdenacao dos Programas de Pos-Graduacao em Egenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2015-07-01
Nuclear explosions are usually described in terms of its total yield and associated shock wave, thermal radiation and nuclear radiation effects. The nuclear radiation produced in such events has several components, consisting mainly of alpha and beta particles, neutrinos, X-rays, neutrons and gamma rays. For practical purposes, the radiation from a nuclear explosion is divided into {sup i}nitial nuclear radiation{sup ,} referring to what is issued within one minute after the detonation, and 'residual nuclear radiation' covering everything else. The initial nuclear radiation can also be split between 'instantaneous or 'prompt' radiation, which involves neutrons and gamma rays from fission and from interactions between neutrons and nuclei of surrounding materials, and 'delayed' radiation, comprising emissions from the decay of fission products and from interactions of neutrons with nuclei of the air. This work aims at presenting isodose curves calculations at ground level by Monte Carlo simulation, allowing risk assessment and consequences modeling in radiation protection context. The isodose curves are related to neutrons produced by the prompt nuclear radiation from a hypothetical nuclear explosion with a total yield of 20 KT. Neutron fluency and emission spectrum were based on data available in the literature. Doses were calculated in the form of ambient dose equivalent due to neutrons H*(10){sub n}{sup -}. (author)
Benchmark calculations with simple phantom for neutron dosimetry (2)
International Nuclear Information System (INIS)
Yukio, Sakamoto; Shuichi, Tsuda; Tatsuhiko, Sato; Nobuaki, Yoshizawa; Hideo, Hirayama
2004-01-01
Benchmark calculations for high-energy neutron dosimetry were undertaken after SATIF-5. Energy deposition in a cylindrical phantom with 100 cm radius and 30 cm depth was calculated for the irradiation of neutrons from 100 MeV to 10 GeV. Using the ICRU four-element loft tissue phantom and four single-element (hydrogen, carbon, nitrogen and oxygen) phantoms, the depth distributions of deposition energy and those total at the central region of phantoms within l cm radius and at the whole region of phantoms within 100 cm radius were calculated. The calculated results of FLUKA, MCNPX, MARS, HETC-3STEP and NMTC/JAM codes were compared. It was found that FLUKA, MARS and NMTC/JAM showed almost the same results. For the high-energy neutron incident, the MCNP-X results showed the largest ones in the total deposition energy and the HETC-3STEP results show'ed smallest ones. (author)
Calculation of neutron albedo from laminated semiinfinite media
International Nuclear Information System (INIS)
Dobrynin, Yu.L.; Mikaehlyan, L.A.; Skorokhvatov, M.D.
1978-01-01
A version of a laminated neutron detector with increased efficiency for recording external neutron fluxes by gamma-quanta from neutron capture is considered. The detector comprises two zones. The first zone constitutes an absorbent layer (europium oxide) 0.5 cm thick, and the second one is a moderator (water with gadolinium salt at the concentration of 0.8 g/l). Mono-energetic neutrons fall normally onto the detector surface. Neutron energy varied from 0.1 eV to MeV. The results of calculations of the integral numerical current albedo (INCA) of neutrons by the Monte Carlo method are presented. The INCA dependences on neutron energy are obtained for one moderator with different gadolinium contents; for the absorbent with the moderator containing and lacking the gadolinium. The resultant dependences are indicative of preferential capture of neutrons by the gadolinium in the moderator, this being more desirable for recording neutrons in the (n, γ) reaction
Equivalent-spherical-shield neutron dose calculations
International Nuclear Information System (INIS)
Russell, G.J.; Robinson, H.
1988-01-01
Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab
Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor
International Nuclear Information System (INIS)
Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.
2008-01-01
Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)
Calculation of the power factor using the neutron diffusion hybrid equation
International Nuclear Information System (INIS)
Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino
2013-01-01
Highlights: ► A neutron diffusion hybrid equation with an external neutron source was used. ► Nodal expansion method to obtain the neutron flux was used. ► Nuclear power factors in each fuel element in the reactor core were calculated. ► The results obtained were very accurate. -- Abstract: In this paper, we used a neutron diffusion hybrid equation with an external neutron source to calculate nuclear power factors in each fuel element in the reactor core. We used the nodal expansion method to obtain the neutron flux for a given control rods bank position. The results were compared with results obtained for eigenvalue problem near criticality condition and fixed source problem during the start-up of the reactor, where external neutron sources are extremely important for the stabilization of external neutron detectors.
Neutron energy spectra calculations in the low power research reactor
International Nuclear Information System (INIS)
Omar, H.; Khattab, K.; Ghazi, N.
2011-01-01
The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)
Neutron shielding for a 252 Cf source
International Nuclear Information System (INIS)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente
2006-01-01
To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source. During calculations a detailed model for the 252 Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252 Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)
Energy Technology Data Exchange (ETDEWEB)
Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)
2016-11-15
Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.
About reliability of WWER pressure vessel neutron fluence calculation
Energy Technology Data Exchange (ETDEWEB)
Belousov, S; Ilieva, K; Antonov, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika
1996-12-31
This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs.
About reliability of WWER pressure vessel neutron fluence calculation
International Nuclear Information System (INIS)
Belousov, S.; Ilieva, K.; Antonov, S.
1995-01-01
This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs
Calculation of neutron importance function in fissionable assemblies using Monte Carlo method
International Nuclear Information System (INIS)
Feghhi, S. A. H.; Afarideh, H.; Shahriari, M.
2007-01-01
The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor
Finite volume thermal-hydraulics and neutronics coupled calculations - 15300
International Nuclear Information System (INIS)
Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.
2015-01-01
The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly
Filtered thermal neutron captured cross-sections measurements and decay heat calculations
International Nuclear Information System (INIS)
Son, Pham Ngoc; Tan, Vuong Huu
2014-01-01
Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)
Neutronic calculations in support of the design of the ITER High Resolution Neutron Spectrometer
International Nuclear Information System (INIS)
Moro, F.; Esposito, B.; Marocco, D.; Villari, R.; Petrizzi, L.; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Dapena, M.
2011-01-01
This paper presents the results of neutronic calculations performed to address important issues related to the optimization of the ITER HRNS (High resolution Neutron Spectrometer) design, in particular concerning the definition of the collimator and the choice of the detector system. The calculations have been carried out using the MCNP5 Monte Carlo code in a full 3-D geometry. The HRNS collimation system has been included in the latest MCNP ITER 40 o model (Alite-4). The ITER scenario 2 reference DT plasma fusion neutron source peaked at 14.1 MeV with Gaussian energy distribution has been used. Neutron fluxes and energy spectra (>1 MeV) have been evaluated at different positions along the HRNS collimator and at the detector location. The noise-to-signal ratio (i.e. the ratio of collided to uncollided neutrons), the breakdown of the collided spectrum into its components, the dependency on the first wall aperture and the gamma-ray spectra at the detector position have also been analyzed. The impact of the results on the design of the HRNS diagnostic system is discussed.
CAP vessel monitoring. Programme, measurement and neutron calculation
International Nuclear Information System (INIS)
Farrugia, J.M.; Nimal, J.C.; Totth, B.; Lloret, R.; Perdreau, R.
1982-03-01
Starting with the design of the CAP (Prototype Advanced NSSS), a programme for pressure vessel monitoring has been prepared, including dosimetry. The dosimetry programme encompasses activation dosimeters (Cu, Nb, Co) and fission dosimeters ( 237 Np, 238 U) installed either inside the pressure vessel with the monitoring test-samples, or in a counting tube outside the pressure vessel. In the first place, a description of the method for neutronic calculation is given; such calculations use the codes ANISN and MERCURE 4 allowing assessment of the neutron spectrum seen by the detectors and the related reaction coefficient. This is followed by a description of the instrumentation. The initial dosimetry results available after the initial operating cycles concur with calculations [fr
International Nuclear Information System (INIS)
Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.
2010-01-01
In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
International Nuclear Information System (INIS)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.
2003-01-01
There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)
International Nuclear Information System (INIS)
Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei
1999-01-01
In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments
Calculation of the importance-weighted neutron generation time using MCNIC method
International Nuclear Information System (INIS)
Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.
2008-01-01
In advanced nuclear power systems, such as ADS, the need for reliable kinetics parameters is of considerable importance because of the lower value for β eff due to the large amount of transuranic elements loaded in the core of those systems. All reactor kinetic parameters are weighted quantities. In other words each neutron with a given position and energy is weighted with its importance. Neutron generation time as an important kinetic parameter, in all nuclear power systems has a significant role in the analysis of fast transients. The difference between non-weighted neutron generation time; Λ; standard in most Monte Carlo codes; and the weighted one Λ + can be quite significant depending on the type of the system. In previous work, based on the physical concept of neutron importance, a new method; MCNIC; using the MCNP code has been introduced for the calculation of neutron importance in fissionable assemblies for all criticality states. In the present work the applicability of MCNIC method has been extended for the calculation of the importance-weighted neutron generation time. The influence of reflector thickness on importance-weighted neutron generation time has been investigated by the development of an auxiliary code, IWLA, for a hypothetic assembly. The results of these calculations were compared with the non-weighted neutron generation times calculated using the Monte Carlo code MCNP. The difference between the importance-weighted and non-weighted quantity is more significant in a reflected system and increases with reflector thickness
Cosmic-ray neutron simulations and measurements in Taiwan
International Nuclear Information System (INIS)
Chen, Wei-Lin; Jiang, Shiang-Huei; Sheu, Rong-Jiun
2014-01-01
This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm -2 in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 x 10 -3 n cm -2 s -1 , was suggested. The cosmic-ray neutron background in Taiwan was studied using the FLUKA simulations and field measurements. A new measurement was performed using a car-mounted high-efficiency neutron detector, re-coding real-time neutron counting rates from sea level up to 3275 m. The attenuation of cosmic-ray neutrons in the lower atmosphere exhibited an effective attenuation length of 163 g cm -2 . The calculated neutron counting rates over predicted the measurements by ∼32 %, which leaded to a correction factor for the FLUKA-calculated cosmic-ray neutrons in the lower atmosphere in Taiwan. In addition, a previous measurement regarding neutron spectrum variation near the air/ground and air/water interfaces was re-evaluated. The results showed that the
Neutronic calculations for a subcritical system with external source
International Nuclear Information System (INIS)
Cintas, A; Lopasso, E.M; Marquez Damian, J. I
2006-01-01
We present a neutronic study on an A D S, systems capable of transmute minor actinides and fission products in order to reduce their radiotoxicity and mean-life.We compare neutronic parameters obtained with Scale/Tort and M C N P modelling a sub-critical system with source from a N E A Benchmark.Due to lack of nuclear data at the temperature of the system, we perform calculations at available temperature of libraries (300 K); to compensate the reactivity insertion due to the temperature change we reduce the size of the fuel zone in order to get a sub-critical system that allow u s to evaluate neutronic parameters of the system with source.We have found that the numerical results (neutron spectrum, neutron flux distributions and other neutronic parameters) are in agreement with the M C N P and with those of the benchmark participants even though the geometric models used are not exactly the same. We conclude that with the real temperature cross sections, the calculation scheme developed (Scale/Tort and M C N P) will give reliable results in A D S evaluations [es
Calculation of Spectra of Neutrons and Charged Particles Produced in a Target of a Neutron Generator
Gaganov, V. V.
2017-12-01
An algorithm for calculating the spectra of neutrons and associated charged particles produced in the target of a neutron generator is detailed. The products of four nuclear reactions 3H( d, n)4He, 2H( d, n)3He, 2H( d, p)3H, and 3He( d, p)4He are analyzed. The results of calculations are presented in the form of neutron spectra for several emission angles and spectra of associated charged particles emitted at an angle of 180° for a deuteron initial energy of 0.13 MeV.
Factors affecting neutron measurements and calculations. Part F. Water content in granite
International Nuclear Information System (INIS)
Iwatani, Kazuo; Hasai, Hiromi; Shizuma, Kiyoshi; Hoshi, Masaharu; Endo, Satoru; Oka, Takamitsu; Imanaka, Tetsuji
2005-01-01
As part of the DS02 studies to reevaluate neutrons from the atomic bomb, we cored rock samples from a pillar of Motoyasu Bridge, located at a distance of 128 m from the hypocenter in Hiroshima, and measured the depth profile of induced 152 Eu radioactivity in the rock (Hasai et al. 1987). By use of the MCNP neutron transport calculation code, the depth profile of 152 Eu in the rock was calculated, assuming a neutron distribution at the given location around the pillar based on the DS86 calculations. The depth profile was then compared with the distribution of measurements (Endo et al. 1999). For the calculation, it is necessary to know the major components of the rock. It is also necessary to estimate the water content correctly, since the cross section of hydrogen-neutron reactions is large, and neutron moderation effects of hydrogen are significant. For this purpose, the basic characteristics of water content in rock were studied, based on a few characteristic experiments to estimate the water content, which was then used in neutron transport calculations. The following describes our concepts and methods. (author)
Neutron shielding for a {sup 252} Cf source
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Manzanares A, E.; 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); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com
2006-07-01
To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source. During calculations a detailed model for the {sup 252}Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare {sup 252}Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)
Energy Technology Data Exchange (ETDEWEB)
Guyonvarh, M; Lecomte, P; Le Meur, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1967-07-01
It is necessary to know, in irradiation loops, the thermal neutron flux after the irradiation device has been introduced and without being obliged to wait for the discharge of this device. In order to measure the flux and to control it continuously, one possible method is to place in the flux a coiled steel tube through which air passes. By measuring the activity of argon 41, and with a knowledge of the flow rate and the temperature of the air, it is possible to calculate the flux. An air-circulation flux controller is described and the relationship between the flux and the count rate is established The accuracy of an absolute measurement is about 14 per cent; that of a relative measurement is about 3 per cent. The measurement can be carried out equally well whether the reactor is operating at maximum or at low power. The measurement range goes from 10{sup 9} to lO{sup 15} n.cm{sup -2}.sec{sup -1}, and it would be possible after a few modifications to measure fluxes between 10{sup 5} and 10{sup 15} n.cm{sup -2}.sec{sup -1}. Finally, the method is very safe to operate: there is little risk of irradiation because of the low specific activity of the argon-41 formed, and no risk of contamination because the decay product of argon-41 is stable. This method, which is now being used in loops, is thus very practical. (authors) [French] Sur des boucles d'irradiation il est necessaire de connaitre le flux de neutrons thermiques apres mise en place du dispositif d'irradiation et sans etre oblige d'attendre le detournement de ce dispositif. Pour mesurer le flux et le controler en permanence, une methode consiste a placer sous flux un serpentin en acier dans lequel on fait circuler de l'air. La mesure d'activite d'argon 41 permet de calculer le flux, connaissant le debit et la temperature de l'air. Un controleur de flux par circulation d'air est decrit et la relation entre le flux et le taux de comptage est etablie. La precision d'une mesure absolue est de l'ordre de 14 pour
Experiments and calculations on neutron streaming through bent ducts
Energy Technology Data Exchange (ETDEWEB)
Kloosterman, J.L.; Hoogenboom, J.E. (Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.); Zsolnay, E.M.
1993-07-01
Neutron spectra in a cylindrical straight duct and in bent ducts with angles of 30deg, 60deg and 90deg have been measured by the multiple foil activation and thermoluminescence dosimetry methods. Two-dimensional discrete ordinates and three-dimensional Monte Carlo calculations are executed, and the results are compared with the measurements. The flow rate at the duct entrance calculated by the DOT3.5 code is underestimated by approximately 30 %, due to a conversion of the core and reflector geometry from XY to RZ geometry. The fast neutron flux in the ducts is underestimated by 20 % by the MORSE-SGC/S code due to a too coarse angular mesh of the source, which does not properly represent the actual angular distribution of the fast flux, which is highly peaked forwardly into the ducts. The thermal neutron flux was over-estimated by the Monte Carlo calculation. A method is proposed to calculate the angular distribution of the flow rate at the duct entrance and to calculate the source strength and the angular distribution of the flow rate at the entrance of the second leg of the duct. The results are compared with those of the transport calculations. Generally, the agreement is quite satisfactory. (author).
Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4
International Nuclear Information System (INIS)
Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R.
2014-01-01
Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H p (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm 3 , composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm 2 ). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)
Cross-section calculations for neutron-induced reactions up to 50 MeV
International Nuclear Information System (INIS)
Yamamuro, Nobuhiro.
1996-01-01
In the field of accelerator development, medium-energy reaction cross-section data for structural materials of accelerator and shielding components are required, especially for radiation protection purposes. For a d + Li stripping reaction neutron source used in materials research, neutron reaction cross sections up to 50 MeV are necessary for the design study of neutron irradiation facilities. The current version of SINCROS-II is able to calculate neutron and proton-induced reaction cross sections up to ∼ 50 MeV with some modifications and extensions of the cross-section calculation code. The production of isotopes when structural materials and other materials are bombarded with neutrons or protons is calculated using a revised code in the SINCROS-II system. The parameters used in the cross-section calculations are mainly examined with proton-induced reactions because the experimental data for neutrons above 20 MeV are rare. The status of medium mass nuclide evaluations for aluminum, silicon, chromium, manganese, and copper is presented. These data are useful to estimate the radiation and transmutation of nuclei in the materials
Development of concept and neutronic calculation method for large LMFBR core
International Nuclear Information System (INIS)
Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.
1991-01-01
Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)
Hyperfine 3D neutronic calculations in CANDU supercells
International Nuclear Information System (INIS)
Balaceanu, V.; Aioanei, L.; Pavelescu, M.
2010-01-01
For an accurate evaluation of the fuel performances, it is very important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle. According this issue, in our Institute, a multigroup calculation methodology named WIMS-PIJXYZ was especially developed for estimating the local neutronic parameters in CANDU cell/supercells. The objective of this paper is to present this calculation methodology and to use it in performing some hyperfine neutronic calculations in CANDU type supercells. More exactly, after a short description for the WIMS-PIJXYZ methodology, the end effect for some CANDU fuel bundles is estimated. The WIMS-PIJXYZ methodology is based on WIMS and PIJXYZ transport codes. WIMS is a standard lattice-cell code and it is used for generating the multigroup macroscopic cross sections for the materials in the fuel cells. For obtaining the flux and power distributions in CANDU fuel bundles the PIJXYZ code is used. This code is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The end effect consists in the increasing of the thermal neutron flux in the end region and the increasing of power in the end of the fuel rod. The region separating the CANDU fuel in two adjoining bundles in a channel is called the 'end region' and the end of the last pellet in the fuel stack adjacent to the end region is called the 'fuel end'. The end effect appears because the end region of the bundle is made up of coolant and Zircaloy-4, a very low neutron absorption material. To estimate the end effect, the flux peaking factors and the power peaking factors are calculated. It was taken in consideration CANDU Standard (Natural Uranium, with 37 elements) fuel bundles. In the end of the paper, the results obtained by WIMS-PIJXYZ methodology with the similar LEGENTR results are compared. The comparative analysis shows a good agreement. (authors)
RA-0 reactor. New neutronic calculations
International Nuclear Information System (INIS)
Rumis, D.; Leszczynski, F.
1990-01-01
An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author) [es
Calculation of neutron spectra produced in neutron generator target: Code testing.
Gaganov, V V
2018-03-01
DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
West, D.
1985-07-01
The application of neutron coincidence counting to the assay of special nuclear material involves a major correction for neutron multiplication. The correction commonly used at present requires an accurate knowledge of the intensity ratio of neutrons from (α,n) reactions to those from spontaneous fission. This paper covers various factors, which need to be evaluated in order to assess their importance, in the calculation of (α,n) neutron production using measured thick target yields. They include: accuracy of (α,n) thick target yield measurements; errors introduced by deriving yields in compounds from the measured yields in the constituents and vice-versa; the likely effect of neglecting the difference of α-particle stopping power between Pu and U on the calculated neutron yield from mixed oxide fuel pellets; the intensity of neutrons produced from 1 to 2% of Al used to alloy plutonium metal; the intensity of neutrons produced in Al, used as canning material, from α-particles escaping from the surface layers of oxide or metal fuel; and neutron production from oxygen in the air spaces of powdered PuO 2 prior to sintering. (author)
Neutronic calculations of PARR-1 cores using LEU silicide fuel
International Nuclear Information System (INIS)
Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.
1991-08-01
Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)
DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program
International Nuclear Information System (INIS)
Courtney, J. C.
1987-01-01
1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected
Fusion neutron yield and flux calculation on HT-7 superconducting tokamak
International Nuclear Information System (INIS)
Fu Yanzhang; Zhu Yubao; Chen Juequan
2006-01-01
Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)
MATERIAL COMPOSITIONS AND NUMBER DENSITIES FOR NEUTRONICS CALCULATIONS
International Nuclear Information System (INIS)
D. A. Thomas
1996-01-01
The purpose of this analysis is to calculate the number densities and isotopic weight percentages of the standard materials to be used in the neutronics (criticality and radiation shielding) evaluations by the Waste Package Development Department. The objective of this analysis is to provide material number density information which can be referenced by future neutronics design analyses, such as for those supporting the Conceptual Design Report
Benchmark neutron porosity log calculations
International Nuclear Information System (INIS)
Little, R.C.; Michael, M.; Verghese, K.; Gardner, R.P.
1989-01-01
Calculations have been made for a benchmark neutron porosity log problem with the general purpose Monte Carlo code MCNP and the specific purpose Monte Carlo code McDNL. For accuracy and timing comparison purposes the CRAY XMP and MicroVax II computers have been used with these codes. The CRAY has been used for an analog version of the MCNP code while the MicroVax II has been used for the optimized variance reduction versions of both codes. Results indicate that the two codes give the same results within calculated standard deviations. Comparisons are given and discussed for accuracy (precision) and computation times for the two codes
International Nuclear Information System (INIS)
Kocherov, N.P.
1996-01-01
The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs
Evaluation and calculation of neutron transactinide cross-sections
International Nuclear Information System (INIS)
Konshin, V.A.
1980-01-01
This paper reviews the state of the art of nuclear theory and its application to the evaluation and calculation of neutron reaction cross sections of transactinium isotopes. In particular, the paper describes the current evaluation of the total files of neutron reaction data for 240 Pu and 241 Pu in the energy range between 10 -5 eV and 15 MeV based on a thorough analysis of available experimental data and on the use of modern theoretical concepts, and the work in progress on the evaluation of the total neutron reaction data file for 242 Pu and 241 Am. (author)
International fusion materials irradiation facility and neutronic calculations for its test modules
International Nuclear Information System (INIS)
Sokcic-Kostic, M.
1997-01-01
The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs
Comparison of neutron transport calculations with NRC test results
International Nuclear Information System (INIS)
Koban, J.; Hofmann, W.
1981-02-01
For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de
International Nuclear Information System (INIS)
Bussac, J.; Reuss, P.
1985-01-01
This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr
Neutronic investigation and activation calculation for CFETR HCCB blankets
Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI
2017-12-01
The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.
Calculation of neutron die-away times in a large-vehicle portal monitor
International Nuclear Information System (INIS)
Lillie, R.A.; Santoro, R.T.; Alsmiller, R.G. Jr.
1980-05-01
Monte Carlo methods have been used to calculate neutron die-away times in a large-vehicle portal monitor. These calculations were performed to investigate the adequacy of using neutron die-away time measurements to detect the clandestine movement of shielded nuclear materials. The geometry consisted of a large tunnel lined with He 3 proportional counters. The time behavior of the (n,p) capture reaction in these counters was calculated when the tunnel contained a number of different tractor-trailer load configurations. Neutron die-away times obtained from weighted least squares fits to these data were compared. The change in neutron die-away time due to the replacement of cargo in a fully loaded truck with a spherical shell containing 240 kg of borated polyethylene was calculated to be less than 3%. This result together with the overall behavior of neutron die-away time versus mass inside the tunnel strongly suggested that measurements of this type will not provide a reliable means of detecting shielded nuclear materials in a large vehicle. 5 figures, 4 tables
Calculation of quality factor for monoenergetic neutrons - in accordance with ICRU Report 40
International Nuclear Information System (INIS)
Bregadze, Y.I.; Maslyaev, P.F.; Nurlibaev, K.N.
1988-01-01
The quality factors for heavy particle dose in tissue from first interactions of monoenergetic neutrons calculated directly from the new quality factor determination given in ICRU Report 40 are presented as a function of neutron energy. The results of the calculation are compared with previously published calculations based on old concepts. For neutron energies from 100 keV up to 1 MeV these differences are about a factor of two. (author)
Calculation of efficiency of high-energy neutron detection by plastic scintillators
International Nuclear Information System (INIS)
Telegin, Yu.N.
1977-01-01
A computer was used to calculate neutron (5-30O MeV) registration effeciencies with plastic scintillators 2,5,10, 20,30,40 and 50 cm thick. The results are shown in the form of tables. The contributions to efficiency of various processes have been analysed. The calculation results may be used in planning experiments with neutron counters
Program system for calculating streaming neutron radiation field in reactor cavity
International Nuclear Information System (INIS)
He Zhongliang; Zhao Shu.
1986-01-01
The A23 neutron albedo data base based on Monte Carlo method well agrees with SAIL albedo data base. RSCAM program system, using Monte Carlo method with albedo approach, is used to calculate streaming neutron radiation field in reactor cavity and containment operating hall. The dose rate distributions calculated with RSCAM in square concrete duct well agree with experiments
Delayed neutron spectra and their uncertainties in fission product summation calculations
Energy Technology Data Exchange (ETDEWEB)
Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)
1997-03-01
Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)
Sensitivity Calculation of Vanadium Self-Powered Neutron Detector
International Nuclear Information System (INIS)
Cha, Kyoon Ho
2011-01-01
Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the rhodium SPND has been used in many nuclear power plants. The lifetime of rhodium is too short (about 3∼5 years) to operate the nuclear power plant economically. The vanadium (V) SPND is also primarily sensitive to neutrons like rhodium, but is a somewhat slower reaction time as that of a rhodium SPND. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. For this reason, a vanadium SPND has been being developed to replace the rhodium SPND which is used in OPR1000. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina (Al 2 O 3 ) insulator with a cylindrical geometry. An MCNP-X code was used to simulate some factors (neutron self shielding factor and electron escape probability from the emitter) necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND
Calculation of neutron-induced single-event upset cross sections for semiconductor memory devices
International Nuclear Information System (INIS)
Ikeuchi, Taketo; Watanabe, Yukinobu; Nakashima, Hideki; Sun, Weili
2001-01-01
Neutron-induced single-event upset (SEU) cross sections for semiconductor memory devices are calculated by the Burst Generation Rate (BGR) method using LA150 data and QMD calculation in the neutron energy range between 20 MeV and 10 GeV. The calculated results are compared with the measured SEU cross sections for energies up to 160 MeV, and the validity of the calculation method and the nuclear data used is verified. The kind of reaction products and the neutron energy range that have the most effect on SEU are discussed. (author)
International Nuclear Information System (INIS)
Khelifi, R.; Idiri, Z.; Bode, P.
2002-01-01
The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA
Energy Technology Data Exchange (ETDEWEB)
Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)
2001-07-01
Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)
International Nuclear Information System (INIS)
Vermeeren, L.
2001-01-01
Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)
Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko
2017-09-01
In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.
Neutronic parameters calculations of a CANDU reactor
International Nuclear Information System (INIS)
Zamonsky, G.
1991-01-01
Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author) [es
Neutronic calculations for Angra-1 steam line break accident
International Nuclear Information System (INIS)
Ponzoni Filho, Pedro; Sato, Sadakatu
2000-01-01
The reduction of boron concentration in the Boron Injection Tank (BIT), to the room temperature solubility level, makes necessary a reanalysis of the steam line break accident of Angra 1 NPP. This paper describes the neutronic calculation related to this reanalysis. The main steps of the work were: review of reactivity parameters used in the accident simulation; search of xenon profiles that cause the most severe core power distribution; calculation of hot channel factors and other neutronic parameters necessary for DNBR determination. The final conclusion, related to the steam line break accident, states the BIT concentration may be reduced to 2000 ppm. (author)
International Nuclear Information System (INIS)
Andrejsek, K.
1977-01-01
The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)
Calculating the energy spectrum of neutrons from tritium target of the NG-150 type generator
International Nuclear Information System (INIS)
Bortash, A.I.; Kuznetsov, V.S.
1987-01-01
Calculation procedure of neutron spectra yielding from the NG-150 generator target chamber with regard to deutron moderation is suggested. Using the suggested procedure, neutron spectra for different escape angles formed in the tritium target are calculated. The spectrum of neutrons scattered in cooling water is calculated. The mean energy of neutrons escaping at the angle of 0 deg equalling 14.5 MeV is obtained
International Nuclear Information System (INIS)
Egbert, Stephen D.
2005-01-01
In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)
Transport calculations for a 14.8 MeV neutron beam in a water phantom
International Nuclear Information System (INIS)
Goetsch, S.J.
1981-01-01
A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented
CALCULATION OF AIR ION REGIME IN THE CASE OF ARTIFICIAL AIR IONIZATION
Directory of Open Access Journals (Sweden)
BILIAIEV M. M.
2015-10-01
Full Text Available Purpose. One of the major tasks in the field of labor protection is providing of the necessary qualitative composition of air in the working areas of office and industrial spaces. In order to maintain the necessary air ion level in the air space premises, the artificial ionization of air is used often in the premises. At present in Ukraine analytical model are used for the calculation of air ion regime in premises, influencing on the formation process of air ions concentration field. An alternative solution is the use of CFD models, developing including the air jets aerodynamics in the premise, the presence of furniture, equipment, transfer of ions under an electric field, and other physical factors, determining intensity and shape of air ions concentration field in the premise. Methodology. Influence of air flow was taken into account in the development of CFD models for calculation of air ion regime in the apartment, caused by operation of ventilation, diffusion, electric field impact, as well as the interaction of different polarity ions with each other, and their interaction with dust particles. The proposed model of calculation of air ion regime in premises based on the use of aerodynamics, electrostatics and mass transfer levels. This model allows operatively to calculate air ions concentration field with the influence of the walls, floor, ceiling and obstacles on the process of air ions dispersion, the specific location of different polarity ions emission and their interaction in the premise and work areas in conditions of artificial air ionization. Results. The calculated data were obtained and on their base could be estimated the concentration of air ion anywhere in the premise with artificial air ionization. Ions concentration field, being calculated using this CFD model, as concentration field isolines is presented. Originality. The results of the air ion regime calculation in the premise are presented, based on numerical 2D CFD model
Thermal neutron dose calculation in synovium membrane for BNCS
International Nuclear Information System (INIS)
Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.
2006-01-01
A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values. (author)
A study of the cosmic-ray neutron field near interfaces
Sheu, R J; Jiang, S H
2002-01-01
This study investigated the characteristics of the cosmic-ray neutron field near air/ground and air/water interfaces with an emphasis on the angular distribution. Two sets of high-efficiency neutron detecting systems were used. The first one, called the Bonner Cylinders, was used for measurements of the energy information. The other one, referred to as the eight-channel neutron detector (8CND), was used to characterize the angular information of the neutron field. The measured results were used to normalize and confirm one-dimensional transport calculations for cosmic-ray neutrons below 20 MeV in the air/ground and air/water media. Annual sea level cosmic-ray neutron doses were then determined based on the obtained characteristics of low-energy cosmic-ray neutrons near interfaces and estimated contribution from high-energy neutrons.
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.)
Contribution to a neutronic calculation scheme for pressurized water reactors
International Nuclear Information System (INIS)
Martin Del Campo, C.
1987-01-01
This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost
Neutron shielding point kernel integral calculation code for personal computer: PKN-pc
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.
1994-07-01
A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR
International Nuclear Information System (INIS)
Kurosawa, M.
2005-01-01
For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.
Kurosawa, Masahiko
2005-01-01
For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.
Energy Technology Data Exchange (ETDEWEB)
Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)
2013-09-15
Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.
International Nuclear Information System (INIS)
Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.
2013-01-01
Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function
Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions
International Nuclear Information System (INIS)
Santiago, Daniela Maiolino Norberto
2011-01-01
In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)
Current trends in methods for neutron diffusion calculations
International Nuclear Information System (INIS)
Adams, C.H.
1977-01-01
Current work and trends in the application of neutron diffusion theory to reactor design and analysis are reviewed. Specific topics covered include finite-difference methods, synthesis methods, nodal calculations, finite-elements and perturbation theory
Neutron transport calculations of some fast critical assemblies
Energy Technology Data Exchange (ETDEWEB)
Martinez-Val Penalosa, J A
1976-07-01
To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.
Neutron transport calculations of some fast critical assemblies
International Nuclear Information System (INIS)
Martinez-Val Penalosa, J. A.
1976-01-01
To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs
International Nuclear Information System (INIS)
Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.
1999-01-01
The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements
Calculation of the thermal neutron flux depression in the loop VISA-1
International Nuclear Information System (INIS)
Martinc, R.
1961-01-01
Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr
Recent advances in neutron tomography
International Nuclear Information System (INIS)
McFarland, E.; Massachusetts Inst. of Technology, Cambridge, MA; Lanza, R.
1993-01-01
Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method
Energy Technology Data Exchange (ETDEWEB)
Koivunoro, Hanna; Seppaelae, Tiina; Uusi-Simola, Jouni; Merimaa, Katja; Savolainen, Sauli [Department of Physics, POB 64, FI-00014 University of Helsinki (Finland); Kotiluoto, Petri; Seren, Tom; Auterinen, Iiro [VTT Technical Research Centre of Finland, Espoo, POB 1000, FI-02044 VTT (Finland); Kortesniemi, Mika, E-mail: hanna.koivunoro@helsinki.f [HUS Helsinki Medical Imaging Center, University of Helsinki, POB 340, FI-00029 HUS (Finland)
2010-06-21
In this paper, the accuracy of dose planning calculations for boron neutron capture therapy (BNCT) of brain and head and neck cancer was studied at the FiR 1 epithermal neutron beam. A cylindrical water phantom and an anthropomorphic head phantom were applied with two beam aperture-to-surface distances (ASD). The calculations using the simulation environment for radiation application (SERA) treatment planning system were compared to neutron activation measurements with Au and Mn foils, photon dose measurements with an ionization chamber and the reference simulations with the MCNP5 code. Photon dose calculations using SERA differ from the ionization chamber measurements by 2-13% (disagreement increased along the depth in the phantom), but are in agreement with the MCNP5 calculations within 2%. The {sup 55}Mn(n,{gamma}) and {sup 197}Au(n,{gamma}) reaction rates calculated using SERA agree within 10% and 8%, respectively, with the measurements and within 5% with the MCNP5 calculations at depths >0.5 cm from the phantom surface. The {sup 55}Mn(n,{gamma}) reaction rate represents the nitrogen and boron depth dose within 1%. Discrepancy in the SERA fast neutron dose calculation (of up to 37%) is corrected if the biased fast neutron dose calculation option is not applied. Reduced voxel cell size ({<=}0.5 cm) improves the SERA calculation accuracy on the phantom surface. Despite the slight overestimation of the epithermal neutrons and underestimation of the thermal neutrons in the beam model, neutron calculation accuracy with the SERA system is sufficient for reliable BNCT treatment planning with the two studied treatment distances. The discrepancy between measured and calculated photon dose remains unsatisfactorily high for depths >6 cm from the phantom surface. Increasing discrepancy along the phantom depth is expected to be caused by the inaccurately determined effective point of the ionization chamber.
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)
Active neutron multiplicity analysis and Monte Carlo calculations
International Nuclear Information System (INIS)
Krick, M.S.; Ensslin, N.; Langner, D.G.; Miller, M.C.; Siebelist, R.; Stewart, J.E.; Ceo, R.N.; May, P.K.; Collins, L.L. Jr
1994-01-01
Active neutron multiplicity measurements of high-enrichment uranium metal and oxide samples have been made at Los Alamos and Y-12. The data from the measurements of standards at Los Alamos were analyzed to obtain values for neutron multiplication and source-sample coupling. These results are compared to equivalent results obtained from Monte Carlo calculations. An approximate relationship between coupling and multiplication is derived and used to correct doubles rates for multiplication and coupling. The utility of singles counting for uranium samples is also examined
Energy Technology Data Exchange (ETDEWEB)
Nilsson, J; Sandlin, R
1964-12-15
An investigation has been carried out concerning the transmission of thermal and fast neutrons in air filled annular ducts through laminated Fe-D{sub 2}O shields. Measurements have been made with annular air gaps of 0.5, 1.0, 1.5 and 2.0 cm, at a duct length of half a meter. The neutron fluxes were determined with a foil activation technique. The thermal flux was theoretically and experimentally divided into three components, a streaming, a leakage and an albedo component. The fast flux was similarly divided into a streaming component and a 'leakage' component. A calculational model to predict the components was then developed and fitted, to the data obtained by experiments. The model reported here for prediction of neutron attenuation in ducted configurations may be applied to straight annular ducts of arbitrary dimensions and material configurations but is especially designed for the problems met with in short ducts.
Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes
International Nuclear Information System (INIS)
Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.
1988-01-01
The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs
Calculated /alpha/-induced thick target neutron yields and spectra, with comparison to measured data
International Nuclear Information System (INIS)
Wilson, W.B.; Bozoian, M.; Perry, R.T.
1988-01-01
One component of the neutron source associated with the decay of actinide nuclides in many environments is due to the interaction of decay /alpha/ particles in (/alpha/,n) reactions on low Z nuclides. Measurements of (/alpha/,n) thick target neutron yields and associated neutron spectra have been made for only a few combinations of /alpha/ energy and target nuclide or mixtures of actinide and target nuclides. Calculations of thick target neutron yields and spectra with the SOURCES code require /alpha/-energy-dependent cross sections for (/alpha/,n) reactions, as well as branching fractions leading to the energetically possible levels of the product nuclides. A library of these data has been accumulated for target nuclides of Z /le/ 15 using that available from measurements and from recent GNASH code calculations. SOURCES, assuming neutrons to be emitted isotopically in the center-of-mass system, uses libraries of /alpha/ stopping cross sections, (/alpha/,n) reaction cross reactions, product nuclide level branching fractions, and actinide decay /alpha/ spectra to calculate thick target (/alpha/,n) yields and neutron spectra for homogeneous combinations of nuclides. The code also calculates the thick target yield and angle intergrated neutron spectrum produced by /alpha/-particle beams on targets of homogeneous mixtures of nuclides. Illustrative calculated results are given and comparisons are made with measured thick target yields and spectra. 50 refs., 1 fig., 2 tabs
One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity
Energy Technology Data Exchange (ETDEWEB)
Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry, E-mail: zzhong@anl.gov, E-mail: alby@anl.gov, E-mail: gohar@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, IL (United States)
2011-07-01
The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β{sub eff} has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β{sub eff} was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β{sub eff}, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have
One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity
International Nuclear Information System (INIS)
Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry
2011-01-01
The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β_e_f_f has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β_e_f_f was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β_e_f_f, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have been
Transport calculation of neutron flux distribution in reflector of PW reactor
International Nuclear Information System (INIS)
Remec, I.
1982-01-01
Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)
International Nuclear Information System (INIS)
Egbert, Stephen D.
2005-01-01
The 63 Ni measurements for fast neutrons in copper samples are compared to the calculated free-in-air 63 Ni neutron activation given in Chapter 3 by use of transmission factors. Transmission factors were calculated to account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the untilted free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. Examples of the application of TF's will be provided in this section. (author)
A computer code for calculating neutron cross-sections from resonance parameter data
International Nuclear Information System (INIS)
Mill, A.J.
1979-08-01
A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)
International Nuclear Information System (INIS)
Humbert, Ph.
2005-01-01
In this paper we consider the probability distribution of neutrons in a multiplying assembly. The problem is studied using a space independent one group neutron point reactor model without delayed neutrons. We recall the generating function methodology and analytical results obtained by G.I. Bell when the c 2 approximation is used and we present numerical solutions in the general case, without this approximation. The neutron source induced distribution is calculated using the single initial neutron distribution which satisfies a master (Kolmogorov backward) equation. This equation is solved using the generating function method. The generating function satisfies a differential equation and the probability distribution is derived by inversion of the generating function. Numerical results are obtained using the same methodology where the generating function is the Fourier transform of the probability distribution. Discrete Fourier transforms are used to calculate the discrete time dependent distributions and continuous Fourier transforms are used to calculate the asymptotic continuous probability distributions. Numerical applications are presented to illustrate the method. (author)
Energy Technology Data Exchange (ETDEWEB)
Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-05-01
The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)
Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4
International Nuclear Information System (INIS)
Tombakoglu, M.; Cecen, Y.
2001-01-01
In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
Khattab, K.; Sulieman, I.
2009-11-01
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)
Shielding calculations for the design of neutron radiography facility around PARR
International Nuclear Information System (INIS)
Ashraf, M.M.; Khan, A.R.
1989-06-01
Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)
Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor
International Nuclear Information System (INIS)
Bozic, M.; Zagar, T.; Ravnik, M.
2001-01-01
The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)
International Nuclear Information System (INIS)
Boulyga, E.G.; Boulyga, S.F.
2000-01-01
A computer program is described, which calculates the induced activities of isotopes after irradiation in a known neutron field, thermal and epithermal neutron fluxes from the measured induced activities and from nuclear data of 2-4 monitor nuclides as well as the element concentrations in samples irradiated together with the monitors. The program was developed for operation in Windows 3.1 (or higher). The application of the program for neutron activation analysis allows to simplify the experimental procedure and to reduce the time. The program was tested by measuring different types of standard reference materials at the FRJ-2 (Research Centre, Juelich, Germany) and Triga (University Mainz, Germany) reactors. Comparison of neutron flux parameters calculated by this program with those calculated by a VAX program developed at the Research Centre, Juelich was done. The results of testing seem to be satisfactory. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
International Nuclear Information System (INIS)
Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C
International Nuclear Information System (INIS)
Suman, H.; Kharita, M. H.; Yousef, S.
2008-02-01
In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)
Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca
International Nuclear Information System (INIS)
Bosevski, T.; Altiparmakov, D.; Marinkovic, N.
1974-01-01
In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr
EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2
Energy Technology Data Exchange (ETDEWEB)
Dahlfors, Marcus [Uppsala Univ. (Sweden). Dept. of Radiation Sciences; Kadi, Yacine [CERN, Geneva (Switzerland). Emerging Energy Technologies
2006-01-15
The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in {sup 129}I, {sup 237}Np and {sup 243}Am samples and of fission reaction rates in {sup 235}U, {sup 237}Np and {sup 243}Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations.
Calculation of neutron shielding for a real loaded C-30 cask by code DORT
International Nuclear Information System (INIS)
Lacina, J.
1999-01-01
Measured neutron dose rates of real loaded C-30 casks for WWER spent fuel assemblies are compared with calculated values in the frame of benchmark calculation task. The part of this benchmark task concerning neutron shielding was calculated. Neutron sources values were taken from data presented by V. Chrapciak during the eighth symposium Atomic Energy Research, Bystrice pod Perstejnem in 1998 and the data about cask from the article of the same author from the Atomic Energy Research working group E meeting at Stolpen in 1998. (Author)
The problem of resonance self-shielding effect in neutron multigroup calculations
International Nuclear Information System (INIS)
Wang Qingming; Huang Jinghua
1991-01-01
It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations
International Nuclear Information System (INIS)
Artem’ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.
2016-01-01
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10 -7 –10 -3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder
Artem'ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.
2016-01-01
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2-5 nm and for neutron energies 3 × 10-7-10-3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.
Energy Technology Data Exchange (ETDEWEB)
Artem’ev, V. A., E-mail: niitm@inbox.ru [Research Institute of Materials Technology (Russian Federation); Nezvanov, A. Yu. [Moscow State Industrial University (Russian Federation); Nesvizhevsky, V. V. [Institut Max von Laue—Paul Langevin (France)
2016-01-15
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10{sup -7}–10{sup -3} eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.
Model calculations of excitation functions of neutron-induced reactions on Rh
International Nuclear Information System (INIS)
Strohmaier, Brigitte
1995-01-01
Cross sections of neutron-induced reactions on 103 Rh have been calculated by means of the statistical model and the coupled-channels optical model for incident-neutron energies up to 30 MeV. The incentive for this study was a new measurement of the 103 Rh(n, n') 103m Rh cross section which will - together with the present calculations -enter into a dosimetry-reaction evaluation. The validation of the model parameters relied on nuclear-structure data as far as possible. (author)
Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV
Martinkovic, J
2005-01-01
Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.
Monte Carlo calculations for intermediate-energy standard neutron field
International Nuclear Information System (INIS)
Joneja, O.P.; Subbukutty, K.; Iyengar, S.B.D.; Navalkar, M.P.
Intermediate-Energy Standard Neutron Field (ISNF) which produces a well characterised spectrum in the energy range of interest for fast reactors including breeders, has been set up at NBS using thin enriched 235 U fission sources. A proposal has been made for setting up a similar facility at BARC using however, easily available natural U instead of enriched U sources, to start with. In order to simulate the neutronics of such a facility Monte Carlo method of calculations has been adopted and developed. The results of these calculations have been compared with those of NBS and it is found that there may be a maximum difference of 10% in spectrum characteristics for the two cases of using thick and thin fission sources. (K.B.)
Optical model calculation of neutron-nucleus scattering cross sections
International Nuclear Information System (INIS)
Smith, M.E.; Camarda, H.S.
1980-01-01
A program to calculate the total, elastic, reaction, and differential cross section of a neutron interacting with a nucleus is described. The interaction between the neutron and the nucleus is represented by a spherically symmetric complex potential that includes spin-orbit coupling. This optical model problem is solved numerically, and is treated with the partial-wave formalism of scattering theory. The necessary scattering theory required to solve this problem is briefly stated. Then, the numerical methods used to integrate the Schroedinger equation, calculate derivatives, etc., are described, and the results of various programming tests performed are presented. Finally, the program is discussed from a user's point of view, and it is pointed out how and where the program (OPTICAL) can be changed to satisfy particular needs
Directory of Open Access Journals (Sweden)
Avramović Ivana
2007-01-01
Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.
Calculation of fast neutron dose in plastic-coated optical fibers
International Nuclear Information System (INIS)
Siebert, B.R.L.; Henschel, H.
1998-01-01
The dose of fast neutrons in optical fibers with hydrogen-containing coating materials is considerably increased by energetic recoil protons. Their contribution to the dose in a SiO 2 fiber core is calculated by the Monte Carlo method for different fiber geometries and a fiber optic cable. With 14 MeV neutrons the dose in a single fiber is increased by about 21%, whereas in fiber bundles the dose increase can reach about 170%. Maximum dose enhancement in fiber bundles (about 610%) occurs at neutron energies around 5.5 MeV. The dose increase caused by 14 MeV neutrons in the fiber of a typical laboratory cable is about 124%
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi
1987-02-01
Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)
Bourva, L C A
1999-01-01
The general purpose neutron-photon-electron Monte Carlo N-Particle code, MCNP sup T sup M , has been used to simulate the neutronic characteristics of the on-site laboratory passive neutron coincidence counter to be installed, under Euratom Safeguards Directorate supervision, at the Sellafield reprocessing plant in Cumbria, UK. This detector is part of a series of nondestructive assay instruments to be installed for the accurate determination of the plutonium content of nuclear materials. The present work focuses on one aspect of this task, namely, the accurate calculation of the coincidence gate utilisation factor. This parameter is an important term in the interpretative model used to analyse the passive neutron coincidence count data acquired using pulse train deconvolution electronics based on the shift register technique. It accounts for the limited proportion of neutrons detected within the time interval for which the electronics gate is open. The Monte Carlo code MCF, presented in this work, represents...
International Nuclear Information System (INIS)
Yong Sam Chung; Jong Hwa Moon, Young Ju Chung; Seung Yeon Cho; Sang Hun Kang
1999-01-01
The aim of this research was to enhance the use of nuclear analytical techniques for air pollution studies and to study the feasibility of the use of instrumental neutron activation analysis (INAA) as a routine monitoring tool to reveal environmental pollution sources. For the collection of air particulate samples, the Gent stacked filter unit, low volume sampler with Nucleopore membrane filters were used. Trace elements in samples collected at two suburban residential sites, Taejon and Wonju city in the Republic of Korea, were analyzed by INAA. Variations of the elemental concentrations were measured monthly and the enrichment factors were calculated for the fine (< 2 μm EAD) and coarse size (2-10 μm EAD) fractions. The analytical data were treated statistically to estimate the relationship between the two variables, the concentrations of elements and the total suspended particulate matter. The results were used to describe the emission source and their correlation. (author)
Parallel processing of neutron transport in fuel assembly calculation
International Nuclear Information System (INIS)
Song, Jae Seung
1992-02-01
Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's
Connection factor calculation for isotopic neutron flux measurements with foil detectors
International Nuclear Information System (INIS)
Avila L, J.
1987-01-01
Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given
CAREM 25: actual status of the core neutronic design. Calculation line
International Nuclear Information System (INIS)
Lecot, C.A.
1990-01-01
This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author) [es
Calculation of neutron cross sections on iron up to 40 MeV
International Nuclear Information System (INIS)
Arthur, E.D.; Young, P.G.
1980-01-01
The development of high energy d + Li neutron sources for fusion materials radiation damage studies will require neutron cross sections up to 40 MeV. Experimental data above 15 MeV are generally sparse or nonexistent, and reliance must be placed upon nuclear-model calculations to produce the needed cross sections. To satisfy such requirements for the Fusion Materials Irradiation Test Facility (FMIT), neutron cross sections have been calculated for 54 56 Fe between 3 and 40 MeV. These results were joined to the existing ENDF/B-V evaluation below 3 MeV. In this energy range, most neutron reactions can be described using the Hauser-Feshbach statistical model with corrections for preequilibrium and direct-reaction effects. To properly use these models to obtain realistic cross sections, emphasis must be placed upon the determination of suitable input parameters (optical model sets, gamma-ray strength functions, level densities) valid over the energy range of the calculation. To do this, several types of independent data were used to arrive at consistent parameter sets as described
Calculation of contribution of multiple interactions and efficiency of neutron detectors
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Kazakov, L.E.; Kononov, V.N.; Poletaev, E.D.
1986-01-01
Results of calculation of multiple neutron interactions contribution to efficiency of detectors with 6 Li glass and 10 B plate in the energy range of 0.01-1 MeV are given. The calculation was performed by the Monte-Carlo method using BRAND program complex. It is shown that a correction value for multiple neutron interaction in 6 Li glass of 1 mm thickness constitutes 4.5 % at energy of up to 100 keV and at higher energies has a complex energy dependence reaching 25 % at 440 keV
Nuclear structure effects on calculated fast neutron reaction cross sections
International Nuclear Information System (INIS)
Avrigeanu, V.
1992-01-01
The importance of accurate low-lying level schemes for reaction cross section calculation and need for microscopically calculated levels are proved with reference to fast neutron induced reactions in the A = 50 atomic mass range. The uses of the discrete levels both for normalization of phenomenological level density approaches and within Hauser-Feshbach calculations are discussed in this respect. (Author)
The validation of neutron kinetic calculations of CEGB reactors
International Nuclear Information System (INIS)
Emmett, J.C.A.; Hutt, P.K.; Nunn, D.L.; Waterson, R.H.
1982-01-01
Reactor kinetic calculations are required by the CEGB to predict space and time varying neutron fluxes through the course of various hypothesized core transients. These transients arise through flow or reactivity perturbations occurring in a part of the core. A description is given of the results of dual programmes of work undertaken at BNL to validate such calculations. Firstly, analyses have been carried out to establish how data for these calculations should best be derived. Secondly, experimental measurements have been compared against the predictions of such calculations with data derived in the recommended way. (author)
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
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.
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.)
Energy Technology Data Exchange (ETDEWEB)
Villarino, Eduardo; Hergenreder, Daniel [Investigacion Aplicada, INVAP, San Carlos de Bariloche (Argentina)
2000-07-01
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included.
Current status of fast-neutron-capture calculations
International Nuclear Information System (INIS)
Gardner, D.G.
1982-01-01
This work is primarily concerned with the calculation of neutron capture cross sections and capture gamma-ray spectra, in the framework of the Hauser-Feshbach statistical model and for neutrons from the resonance region up to several MeV. An argument is made that, for applied purposes such as constructing evaluated cross-section libraries, nonstatistical capture mechanisms may be completely neglected at low energies and adequately approximated at high energies in a simple way. The use of gamma-ray strength functions to obtain radiation widths is emphasized. Using the reaction 89 Y + n as an example, the problems encountered in trying to construct a case that could be run equivalently on two different nuclear reaction codes are illustrated, and the effects produced by certain parameter variations are discussed
Gamow's calculation of the neutron star's critical mass revised
International Nuclear Information System (INIS)
Ludwig, Hendrik; Ruffini, Remo
2014-01-01
It has at times been indicated that Landau introduced neutron stars in his classic paper of 1932. This is clearly impossible because the discovery of the neutron by Chadwick was submitted more than one month after Landau's work. Therefore, and according to his calculations, what Landau really did was to study white dwarfs, and the critical mass he obtained clearly matched the value derived by Stoner and later by Chandrasekhar. The birth of the concept of a neutron star is still today unclear. Clearly, in 1934, the work of Baade and Zwicky pointed to neutron stars as originating from supernovae. Oppenheimer in 1939 is also well known to have introduced general relativity (GR) in the study of neutron stars. The aim of this note is to point out that the crucial idea for treating the neutron star has been advanced in Newtonian theory by Gamow. However, this pioneering work was plagued by mistakes. The critical mass he should have obtained was 6.9 M, not the one he declared, namely, 1.5 M. Probably, he was taken to this result by the work of Landau on white dwarfs. We revise Gamow's calculation of the critical mass regarding calculational and conceptual aspects and discuss whether it is justified to consider it the first neutron-star critical mass. We compare Gamow's approach to other early and modern approaches to the problem.
International Nuclear Information System (INIS)
Gast, R.C.
1981-08-01
A procedure for defining diffusion coefficients from Monte Carlo calculations that results in suitable ones for use in neutron diffusion theory calculations is not readily obtained. This study provides a survey of the methods used to define diffusion coefficients from deterministic calculations and provides a discussion as to why such traditional methods cannot be used in Monte Carlo. This study further provides the empirical procedure used for defining diffusion coefficients from the RCP01 Monte Carlo program
MCFT: a program for calculating fast and thermal neutron multigroup constants
International Nuclear Information System (INIS)
Yang Shunhai; Sang Xinzeng
1993-01-01
MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825
HEINBE; the calculation program for helium production in beryllium under neutron irradiation
International Nuclear Information System (INIS)
Shimakawa, Satoshi; Ishitsuka, Etsuo; Sato, Minoru
1992-11-01
HEINBE is a program on personal computer for calculating helium production in beryllium under neutron irradiation. The program can also calculate the tritium production in beryllium. Considering many nuclear reactions and their multi-step reactions, helium and tritium productions in beryllium materials irradiated at fusion reactor or fission reactor may be calculated with high accuracy. The calculation method, user's manual, calculated examples and comparison with experimental data were described. This report also describes a neutronics simulation method to generate additional data on swelling of beryllium, 3,000-15,000 appm helium range, for end-of-life of the proposed design for fusion blanket of the ITER. The calculation results indicate that helium production for beryllium sample doped lithium by 50 days irradiation in the fission reactor, such as the JMTR, could be achieved to 2,000-8,000 appm. (author)
International Nuclear Information System (INIS)
Broecker, B.; Clausen, K.; Schneider-Kuehnle, P.; Weinert, M.
1975-01-01
An experiment to measure the radiation transport originating from a 14-MeV neutron source in liquid nitrogen and liquid air is presented. Neutron and gamma-ray spectra were measured with a proton-recoil NE 213 scintillator and with four spherical proportional counters in a tank filled with liquid nitrogen or liquid air. The neutron spectra cover the energy range of 20 keV to 18 MeV. The source-detector separation varies in the liquid medium between 60 and 240 cm. The experimental setup is briefly described and the errors are estimated. (2 tables, 9 figures) (auth)
Calculation of neutron detection efficiency for the thick lithium glass using Monte Carlo method
International Nuclear Information System (INIS)
Tang Guoyou; Bao Shanglian; Li Yulin; Zhong Wenguan
1989-08-01
The neutron detector efficiencies of a NE912 (45mm in diameter, 9.55 mm in thickness) and 2 pieces of ST601 (40mm in diameter, 3 and 10 mm in thickness respectively) lithium glasses have been calculated with a Monte Carlo computer code. The energy range in the calculation is 10 keV to 2.0 MeV. The effect of time delayed caused by neutron multiple scattering in the detectors (prompt neutron detection efficiency) has been considered
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
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
Calculation and applications of the frequency dependent neutron detector response functions
International Nuclear Information System (INIS)
Van Dam, H.; Van Hagen, T.H.J.J. der; Hoogenboom, J.E.; Keijzer, J.
1994-01-01
The theoretical basis is presented for the evaluation of the frequency dependent function that enables to calculate the response of a neutron detector to parametric fluctuations ('noise') or oscillations in reactor core. This function describes the 'field view' of a detector and can be calculated with a static transport code under certain conditions which are discussed. Two applications are presented: the response of an ex-core detector to void fraction fluctuations in a BWR and of both in and ex-core detectors to a rotating neutron absorber near or inside a research reactor core. (authors). 7 refs., 4 figs
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
Calculation of the neutron parameters of fast thermal reactor
International Nuclear Information System (INIS)
Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.
1975-01-01
The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)
Calculations of the relative effectiveness of alanine for neutrons with energies up to 17.1 MeV
International Nuclear Information System (INIS)
Gerstenberg, H.M.; Coyne, J.J.
1990-01-01
The relative effectiveness (RE) of alanine has been calculated for neutrons using the RE of alanine for charged particles. The neutrons interact with one or more of the elements (hydrogen, carbon, nitrogen and oxygen) that compose the alanine. These interactions produce spectra of secondary charged particles consisting of ions of H, D, He, Be, B, C, N and O. From a combination of the calculated secondary charged particle spectra generated by the slowing down neutrons, and the calculated RE of the ions produced, a RE for the neutrons can be obtained. In addition, lineal energy spectra were determined for neutrons with energies up to 17.1 MeV interacting with alanine. An analytical code was used to calculate these spectra for a 1 μm diameter alanine cell surrounded by an alanine medium. For comparison, similar calculations were made for muscle tissue. Finally, the calculated differential RE was folded with dose distributions to obtain RE-weighted distributions for alanine. (author)
Khattab, K; Sulieman, I
2009-04-01
The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.
Diffusion theory model for optimization calculations of cold neutron sources
International Nuclear Information System (INIS)
Azmy, Y.Y.
1987-01-01
Cold neutron sources are becoming increasingly important and common experimental facilities made available at many research reactors around the world due to the high utility of cold neutrons in scattering experiments. The authors describe a simple two-group diffusion model of an infinite slab LD 2 cold source. The simplicity of the model permits to obtain an analytical solution from which one can deduce the reason for the optimum thickness based solely on diffusion-type phenomena. Also, a second more sophisticated model is described and the results compared to a deterministic transport calculation. The good (particularly qualitative) agreement between the results suggests that diffusion theory methods can be used in parametric and optimization studies to avoid the generally more expensive transport calculations
Calculation of background effects on the VESUVIO eV neutron spectrometer
International Nuclear Information System (INIS)
Mayers, J
2011-01-01
The VESUVIO spectrometer at the ISIS pulsed neutron source measures the momentum distribution n(p) of atoms by 'neutron Compton scattering' (NCS). Measurements of n(p) provide a unique window into the quantum behaviour of atomic nuclei in condensed matter systems. The VESUVIO 6 Li-doped neutron detectors at forward scattering angles were replaced in February 2008 by yttrium aluminium perovskite (YAP)-doped γ-ray detectors. This paper compares the performance of the two detection systems. It is shown that the YAP detectors provide a much superior resolution and general performance, but suffer from a sample-dependent gamma background. This report details how this background can be calculated and data corrected. Calculation is compared with data for two different instrument geometries. Corrected and uncorrected data are also compared for the current instrument geometry. Some indications of how the gamma background can be reduced are also given
Calculation of background effects on the VESUVIO eV neutron spectrometer
Mayers, J.
2011-01-01
The VESUVIO spectrometer at the ISIS pulsed neutron source measures the momentum distribution n(p) of atoms by 'neutron Compton scattering' (NCS). Measurements of n(p) provide a unique window into the quantum behaviour of atomic nuclei in condensed matter systems. The VESUVIO 6Li-doped neutron detectors at forward scattering angles were replaced in February 2008 by yttrium aluminium perovskite (YAP)-doped γ-ray detectors. This paper compares the performance of the two detection systems. It is shown that the YAP detectors provide a much superior resolution and general performance, but suffer from a sample-dependent gamma background. This report details how this background can be calculated and data corrected. Calculation is compared with data for two different instrument geometries. Corrected and uncorrected data are also compared for the current instrument geometry. Some indications of how the gamma background can be reduced are also given.
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
Henryson, H. II; Toppel, B.J.; Stenberg, C.G.
1976-06-01
MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
International Nuclear Information System (INIS)
Honeck, H.C.
1984-01-01
1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups
Calculation of the main neutron parameters of the IEA-R1 research reactor
International Nuclear Information System (INIS)
Ojima, Mario Katsuhiko
1977-01-01
The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined
Calculation of prompt neutron spectra for curium isotopes
Energy Technology Data Exchange (ETDEWEB)
Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.
1997-03-01
With the aim of checking the existing evaluations contained in JENDL-3.2 and providing new evaluations based on a methodology proposed by the author, a series of calculations of prompt neutron spectra have been undertaken for curium isotopes. Some of the evaluations in JENDL-3.2 was found to be unphysically hard and should be revised. (author)
Comparison of experimental and calculated neutron emission spectra and angular distributions
International Nuclear Information System (INIS)
Gruppelaar, H.; Akkermans, J.M.
1980-06-01
Experimental and calculated neutron emission spectra and angular distributions have been intercompared for 14.6-MeV neutron-induced reactions. The experimental data, measured by Hermsdorf et al., cover 34 elements in a large mass range. To calculate the differential neutron scattering cross sections a unified model of preequilibrium neutron emission was used, in which the generalized master equation of Mantzouranis et al. was solved with a fast exact matrix method, recently introduced by Akkermans. For the scattering kernel a three-term Legendre polynomial representation was adopted, which was either derived from the differential free nucleon-nucleon scattering cross section or fitted to obtain optimal agreement with the set of experimental data of Hermsdorf et al. The results of the last-mentioned calculation are quite acceptable in view of the fact that only two global parameters have been to describe the angular distributions of all experimental data. The report contains tables and graphs of the calculated Legendre coefficients and graphs of energy-averaged angular distributions for all 34 elements. It is further shown that improvements in the energy and angular distributions could be obtained by means of adjustment of the level-density parameters of the individual residual nuclei. Finally a short discussion is devoted to the problems of fitting angular distributions at backward angles by varying the model parameters or the specification of the initial condition. It is indicated that the so-called preequilibrium phase of the nuclear reaction actually consists of two different stages, the first one generating the forward-peaked angular distributions and the second one showing angular distributions symmetric about 90 0
Impact of neutron resonance treatments on reactor calculation
International Nuclear Information System (INIS)
Leszczynski, F.
1988-01-01
The neutron resonance treatment on reactor calculation is one of the not completely resolved problems of reactor theory. The calculation required on design, fuel management and accident analysis of nuclear reactors contains adjust coefficients and semi-empirical values introduced on the computer codes; these values are obtained comparing calculation results with experimental values and more exact calculation results. This is made when the characteristics of the analyzed system are such that this type of comparisons are possible. The impact that one fixed resonance treatment method have on the final evaluation of physics reactor parameters, reactivity, power distribution, etc., is useful to know. In this work, the differences between calculated parameters with two different methods of resonance treatment in cell calculations are shown. It is concluded that improvements on resonance treatment are necessary for growing the reliability on core calculations results. Finally, possible improvements, easy to implement in current computer codes, are presented. (Author) [es
Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies
International Nuclear Information System (INIS)
Santoro, R.T.; Baker, V.C.; Barnes, J.M.
1976-01-01
Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given
International Nuclear Information System (INIS)
Khattab, K.
2005-03-01
The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)
International Nuclear Information System (INIS)
Khattab, K.
2006-01-01
The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Deeva, V.V.; Prokof'eva, Z.A.
1990-01-01
Analysis is made for the effect of mathematical model accuracy of the system concerned on the calculation results using the BRAND program system. Consideration is given to the impact of the following factors: accuracy of neutron source energy-angular characteristics description, various degrees of system geometry approximation, adequacy of Monte-Carlo method estimation to a real physical neutron detector. The calculation results analysis is made on the basis of the experiments on leakage neutron spectra measurement in spherical lead assemblies with the 14 MeV-neutron source in the centre. 4 refs.; 2 figs.; 10 tabs
Measurement and calculation of fast neutron flux in a zero-energy reactor
International Nuclear Information System (INIS)
Day, D.H.; Fox, W.N.; Hyder, H.R.
1963-05-01
An activation technique for measuring relative fast neutron fluxes is described which has some advantages over the normal method using U238 fission. The technique is based on the formation of Rh 103 after inelastic scattering of neutrons above 100 keV in energy. This isomer decays with a 57.4 minute half-life giving an easily measurable γ-activity. The energy dependence of the inelastic scattering cross-section of Rh 103 is similar to that of the fission cross-section of U 238 thus making the results of direct relevance to reactor calculations. Using the Rh 103 activation technique, measurements have been made of the fast neutron flux distribution in a typical pressure tube heavy water lattice and are compared in this report with theoretical calculations using the MONTE CARLO method. (author)
Linden, P; Dahl, B; Pázsit, I; Por, G
1999-01-01
We have performed laboratory measurements of the neutron flux and its gradient in a static model experiment, similar to a model problem proposed in Pazsit (Ann. Nucl. Energy 24 (1997) 1257). The experimental system consists of a radioactive neutron source located in a water tank. The measurements are performed using a recently developed very small optical fibre detector. The measured values of the flux and its gradient are then used to test the possibility of localising the source. The results show that it is possible to measure the flux on the circumference of a circle and from this calculate the flux gradient vector. Then, by comparison of the measured quantities with corresponding MCNP calculations, both the direction and the distance to the source are found and thus the position of the source can be determined.
International Nuclear Information System (INIS)
Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.
2014-01-01
The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)
International Nuclear Information System (INIS)
Artemov, V.G.; Gusev, V.I.; Zinatullin, R.E.; Karpov, A.S.
2007-01-01
Using modeled WWER cram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases. Numerical modeling was carried out on the basis of SAPFIR 9 5andWWERrogram package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BNAB-78 validated data files. It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results. Based on the results of numerically modeled experiments, delayed neutron parameters providing the best agreement between calculated and measured data were selected and recommended for use in reactor calculations (Authors)
International Nuclear Information System (INIS)
Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.
1985-12-01
A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions
International Nuclear Information System (INIS)
Milosevic, M.
1979-01-01
One-dimensional variational method for cylindrical configuration was applied for calculating group constants, together with effects of elastic slowing down, anisotropic elastic scattering, inelastic scattering, heterogeneous resonance absorption with the aim to include the presence of a number of different isotopes and effects of neutron leakage from the reactor core. Neutron flux shape P 3 and adjoint function are proposed in order to enable calculation of smaller size reactors and inclusion of heterogeneity effects by cell calculations. Microscopic multigroup constants were prepared based on the UKNDL data library. Analytical-numerical approach was applied for solving the equations of the P 3 approximation to obtain neutron flux moments and adjoint functions
International Nuclear Information System (INIS)
Greenwood, L.R.
1983-01-01
This paper is intended as an overview of activities designed to characterize neutron irradiation facilities in terms of neutron flux and energy spectrum and to use these data to calculate atomic displacements, gas production, and transmutation during fusion materials irradiations. A new computerized data file, called DOSFILE, has recently been developed to record dosimetry and damage data from a wide variety of materials test facilities. At present data are included from 20 different irradiations at fast and mixed-spectrum reactors, T(d,n) 14 MeV neutron sources, Be(d,n) broad-spectrum sources, and spallation neutron sources. Each file entry includes activation data, adjusted neutron flux and spectral data, and calculated atomic displacements and gas production. Such data will be used by materials experimenters to determine the exposure of their samples during specific irradiations. This data base will play an important role in correlating property changes between different facilities and, eventually, in predicting materials performance in fusion reactors. All known uncertainties and covariances are listed for each data record and explicit references are given to nuclear decay data and cross sections
Measuring neutron noise induced by travelling air bubbles in a research reactor
International Nuclear Information System (INIS)
Por, G.; Horanyi, S.
1983-05-01
Travelling air bubble experiments carried out in a research reactor confirm an earlier proposed model. The sink structure could be found experimentally in APSD of neutron signals and was used to determine the bubble velocity. The measurements show that neutron detectors measure the velocity of the travelling bubbles, the thermocouples that of the water flow. (author)
Energy Technology Data Exchange (ETDEWEB)
Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)
2014-08-15
By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-10-01
A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)
Talys calculations for evaluation of neutron-induced single-event upset cross sections
Energy Technology Data Exchange (ETDEWEB)
Bourselier, Jean-Christophe
2005-08-15
The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by {sup 28}Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs.
Talys calculations for evaluation of neutron-induced single-event upset cross sections
International Nuclear Information System (INIS)
Bourselier, Jean-Christophe
2005-08-01
The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by 28 Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs
SUNF, Simplified UNF Code, Fast Neutron Calculation by Unified Hauser-Feshbach Theory
International Nuclear Information System (INIS)
Zhang Jingshang
2001-01-01
1 - Description of program or function: The SUNF code is the simplified version of UNF code and is based on the unified Hauser-Feshbach and exciton model. SUNF code has been developed for calculations of fast neutron data for structural materials with neutron energies below 20 MeV. Besides elastic scattering channel, the code may handle decay sequence up to (n,3n) reaction, including 14 reaction channels. The energy spectra can be obtained and the output form is in the ENDF/B-6 format, but in file 5 form. For the ENDF-B-6 output, the incident energies are divided into two types: only cross section calculation; and those including neutron energy spectra. 2 - Methods: Gaussian integration is used for all numerical integration. 3 - Restrictions on the complexity of the problem: The incident energies of neutrons are from 1 KeV to 20 MeV. There are two parameters in this code: incident neutron energies number 'NEL'; and the number of discrete levels of residual nuclei for the first particle emissions 'NLV'. The users can set the values of NEL and NLV according to the storage size of the computer used. The number of discrete levels of residual nuclei for the multi-particle emissions is not greater than 20
Calculating disadvantage factor for fuel taking into account the neutron energy distribution
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1964-01-01
Errors in calculating the disadvantage factor are caused by applying the diffusion approximation and one-group method. This paper describes the method for calculating the fuel disadvantage factor by applying a non-diffusion method taking into account neutron thermalization
International Nuclear Information System (INIS)
Medley, S.S.; Darrow, D.S.; Roquemore, A.L.
2005-01-01
A change in the response of the neutron detectors on the National Spherical Torus Experiment occurred between the 2002-2003 and 2004 experimental run periods. An analysis of this behavior by investigating the neutron diagnostic operating conditions and comparing measured and TRANSP-calculated neutron rates is presented. Also a revised procedure for cross calibration of the neutron scintillator detectors with the fission chamber detectors was implemented that delivers good agreement amongst the measured neutron rates for all neutron detectors and all run periods. For L-mode discharges, the measured and TRANSP-calculated neutron rates now match closely for all run years. For H-mode discharges over the entire 2002-2004 period, the 2FG scintillator and fission chamber measurements match each other but imply a neutron deficit of 11.5% relative to the TRANSP-calculated neutron. The results of this report impose a modification on all of the previously used calibration factors for the entire neutron detector suite over the 2002-2004 period. A tabular summary of the new calibration factors is provided including certified calibration factors for the 2005 run
Neutrons confirmed in Nagasaki and at the Army pulsed radiation Facility: Implications for Hiroshima
International Nuclear Information System (INIS)
Straume, T.; Harris, L.J.; Marchett, A.A.; Egbert, S.D.
1994-01-01
Recent reports have clearly demonstrated that large discrepancies exist between neutron activation measured in Hiroshima and activation calculated using the current dosimetry system DS86. The reports confirmed previous results for cobalt activation in Hiroshoma that suggested problems, and this has spurred a joint U.S.-Japan effort to identify the source(s) of this discrepancy. Here, new results are presented that appear to eliminate both the measurements of neutron activation and the DS86 air-transport calculations are potential sources of the discrepancy in Hiroshima. Computer transport of DS86 fission neutrons through large distances of air was validated using concrete samples from Nagasaki and chloride detectors placed at selected distances from a bare uranium reactor. In both cases accelerator mass spectrometry was used to measure thermal neutron activation via the reaction. 35 Cl(n, γ) 36 Cl (half-life, 301,000 years). Good agreement was observed between measurements of neutron activation and DS86 calculations for Nagasaki, as well as for the reactor experiment. Thus the large discrepancy observed in Hiroshima appears not to be due to uncertainties in air-transport calculations or in the activation measurements; rather, the discrepancy appears to be due to uncertainties associated with the Hiroshima bomb itself. 15 refs., 3 figs., 8 tabs
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.
2007-01-01
In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required
Radiation transport in earth for neutron and gamma ray point sources above an air-ground interface
International Nuclear Information System (INIS)
Lillie, R.A.; Santoro, R.T.
1979-03-01
Two-dimensional discrete ordinates methods were used to calculate the instantaneous dose rate in silicon and neutron and gamma ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 meters) above an air--ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth--concrete model containing 0.9 meters of borated concrete beginning 0.5 meters below the earth's surface to obtain fluence distributions to a depth of 3.0 meters. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon and, in all cases, the secondary gamma ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths greater than 0.6 meter. 4 figures, 4 tables
Radiation transport in earth for neutron and gamma-ray point sources above an air-ground interface
International Nuclear Information System (INIS)
Lillie, R.A.; Santoro, R.T.
1980-01-01
Two-dimensional discrete-ordinates methods have been used to calculate the instantaneous dose rate in silicon and neutron and gamma-ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 m) above an air-ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth-concrete model containing 0.9 m of borated concrete beginning 0.5 m below the earth's surface to obtain fluence distributions to a depth of 3.0 m. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon, and in all cases, the secondary gamma-ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths > 0.6 m
The shell structure effects in neutron cross section calculation by a ...
African Journals Online (AJOL)
The role of the shell structure properties of the nucleus in the calculation of neutron-induced reaction cross-section data based on nuclear reaction theory has been investigated. In this investigation, measured, evaluated and calculated (n.p) reaction cross-section data on la spherical nucleus (i.e. 112Sn) and a deformed ...
Gaganov, V. V.
2017-12-01
The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.
Quality factor calculations for neutron spectra below 4 MeV
International Nuclear Information System (INIS)
Borak, T.B.; Stinchcomb, T.G.
1979-01-01
A method is described for computing the distribution of absorbed dose, D(L), as a function of linear energy transfer, L, for any neutron spectrum with energies below 4 MeV. The results are used to determine the average quality factor for two distinctly different neutron spectra using the ICRP recommended values of the quality factor, Q(L). A comparison is made between the calculations and measurements of D(L) using a spherical tissue equivalent proportional counter. Heavy ion recoil contributions to the average quality factor are examined in detail. (author)
Neutron Flux and Activation Calculations for a High Current Deuteron Accelerator
Coniglio, Angela; Sandri, Sandro
2005-01-01
Neutron analysis of the first Neutral Beam (NB) for the International Thermonuclear Experimental Reactor (ITER) was performed to provide the basis for the study of the following main aspects: personnel safety during normal operation and maintenance, radiation shielding design, transportability of the NB components in the European countries. The first ITER NB is a medium energy light particle accelerator. In the scenario considered for the calculation the accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The average beam current is 13.3 A. To assess neutron transport in the ITER NB structure a mathematical model of the components geometry was implemented into MCNP computer code (MCNP version 4c2. "Monte Carlo N-Particle Transport Code System." RSICC Computer Code Collection. June 2001). The neutron source definition was outlined considering both D-D and D-T neutron production. FISPACT code (R.A. Forrest, FISPACT-2003. EURATOM/UKAEA Fusion, December 2002) was used to assess neutron...
International Nuclear Information System (INIS)
Lee, Y.K.; Hugot, F.X.
2011-01-01
The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)
International Nuclear Information System (INIS)
Wechsler, M.S.; Mansur, L.K.
1996-01-01
Radiation damage in target, container, and window materials for spallation neutron sources is am important factor in the design of target stations for accelerator-driver transmutation technologies. Calculations are described that use the LAHET and SPECTER codes to obtain displacement and helium production rates in tungsten, 316 stainless steel, and Inconel 718, which are major target, container, and window materials, respectively. Results are compared for the three materials, based on neutron spectra for NSNS and ATW spallation neutron sources, where the neutron fluxes are normalized to give the same flux of neutrons of all energies
TORT application in reactor pressure vessel neutron flux calculations
International Nuclear Information System (INIS)
Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.
1994-01-01
The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs
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
Neutron kerma factors, and water equivalence of some tissue substitutes
International Nuclear Information System (INIS)
Singh, V. P.; Badiger, N. M.; Vega C, H. R.
2014-08-01
The kerma factors and kerma relative to air and water of 24 compounds used as tissue substitutes were calculated for neutron energy from 2.53 x 10 -8 up to 29 MeV. The kerma ratio of the tissue substitutes relative to air and water were calculated by the ratio of kerma factors of the tissue substitute to air and water respectively. The water equivalence of the selected tissue substitutes was observed above neutron energies 100 eV. Kerma ratio relative to the air for Poly-vinylidene fluoride and Teflon are found to be nearest to unity in very low energy (up to 1 eV) and above 63 eV respectively. It was found that the natural rubber as a water equivalent tissue substitute compound. The results of the kerma factors in our investigation shows a very good agreement with those published in ICRU-44. We found that at higher neutron energies, the kerma factors and kerma ratios of the selected tissue substitute compounds are approximately same, but differences are large for energies below 100 eV. (Author)
International Nuclear Information System (INIS)
Sasamoto, Nobuo; Kotegawa, Hiroshi
1984-11-01
In order to improve computational efficiency of PALLAS code, an accuracy is estimated on the neutron penetration calculation through a concrete shield, using bunched component nuclides of concrete. The calculated fast neutron flux is observed to depend weakly on how the nuclides are bunched. Contrary to this, the calculated thermal neutron fluxes are strongly dependent on the manner of bunching, mainly due to the fact that iron cross section has exceptionally large negative sensitivity to thermal neutron flux. (author)
Dose distributions in thorax inhomogeneity for fast neutron beam from NIRS cyclotron
International Nuclear Information System (INIS)
Kutsutani-Nakamura, Yuzuru; Furukawa, Shigeo; Iinuma, T.A.; Kawashima, Katsuhiro; Hoshino, Kazuo; Hiraoka, Takeshi; Maruyama, Takashi; Sakashita, Kunio; Tsunemoto, Hiroshi
1990-01-01
The power law tissue-air ratio (TAR) method developed by Batho appears to be practical use for inhomogeneity corrections to the dose calculated in a layered media for photon beam therapy. The validity was examined in applying the modified power law TAR and the isodose shift methods to the dose calculation in thorax tissue inhomogeneity containing the boundary region for fast neutron beam. The neutron beam is produced by bombarding a thick beryllium target with 30 MeV deuterons. Lung phantom was made of granulated tissue equivalent plastic, which resulted in density of 0.30 and 0.60 g/cm 3 . Depth dose distributions for neutron beam were measured in thorax phantom by an air-filled cylindrical ionization chamber with TE plastic wall. The power law TAR method considering TAR of zero depth at boundary was compared with the measured data and a good result was obtained that the calculated dose was within ±3 % against the measured. But the isodose shift method is not so good for dose calculation in thorax tissue inhomogeneity using fast neutron beam. (author)
Benchmark experiments on neutron streaming through JET Torus Hall penetrations
Batistoni, P.; Conroy, S.; Lilley, S.; Naish, J.; Obryk, B.; Popovichev, S.; Stamatelatos, I.; Syme, B.; Vasilopoulou, T.; contributors, JET
2015-05-01
Neutronics experiments are performed at JET for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the neutron fluence through the penetrations of the JET torus hall is measured and compared with calculations to assess the capability of state-of-art numerical tools to correctly predict the radiation streaming in the ITER biological shield penetrations up to large distances from the neutron source, in large and complex geometries. Neutron streaming experiments started in 2012 when several hundreds of very sensitive thermo-luminescence detectors (TLDs), enriched to different levels in 6LiF/7LiF, were used to measure the neutron and gamma dose separately. Lessons learnt from this first experiment led to significant improvements in the experimental arrangements to reduce the effects due to directional neutron source and self-shielding of TLDs. Here we report the results of measurements performed during the 2013-2014 JET campaign. Data from new positions, at further locations in the South West labyrinth and down to the Torus Hall basement through the air duct chimney, were obtained up to about a 40 m distance from the plasma neutron source. In order to avoid interference between TLDs due to self-shielding effects, only TLDs containing natural Lithium and 99.97% 7Li were used. All TLDs were located in the centre of large polyethylene (PE) moderators, with natLi and 7Li crystals evenly arranged within two PE containers, one in horizontal and the other in vertical orientation, to investigate the shadowing effect in the directional neutron field. All TLDs were calibrated in the quantities of air kerma and neutron fluence. This improved experimental arrangement led to reduced statistical spread in the experimental data. The Monte Carlo N-Particle (MCNP) code was used to calculate the air kerma due to neutrons and the neutron fluence at detector positions, using a JET model validated up to the
International Nuclear Information System (INIS)
Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji
1979-01-01
A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)
Neutron noise calculations in a hexagonal geometry and comparison with analytical solutions
International Nuclear Information System (INIS)
Tran, H. N.; Demaziere, C.
2012-01-01
This paper presents the development of a neutronic and kinetic solver for hexagonal geometries. The tool is developed based on the diffusion theory with multi-energy groups and multi-groups of delayed neutron precursors allowing the solutions of forward and adjoint problems of static and dynamic states, and is applicable to both thermal and fast systems with hexagonal geometries. In the dynamic problems, the small stationary fluctuations of macroscopic cross sections are considered as noise sources, and then the induced first order noise is calculated fully in the frequency domain. Numerical algorithms for solving the static and noise equations are implemented with a spatial discretization based on finite differences and a power iterative solution. A coarse mesh finite difference method has been adopted for speeding up the convergence. Since no other numerical tool could calculate frequency-dependent noise in hexagonal geometry, validation calculations have been performed and benchmarked to analytical solutions based on a 2-D homogeneous system with two-energy groups and one-group of delayed neutron precursor, in which point-like perturbations of thermal absorption cross section at central and non-central positions are considered as noise sources. (authors)
Calculation of double energy angle differential neutron albedos for radiation shielding applications
International Nuclear Information System (INIS)
Litaize, O.; Diop, C.M.; Nimal, J.C.
2000-01-01
Void radiation shielding problems can be dealt with albedo concept which is an alternative to the complex bringing into operation of the 'exact' transport method calculations (SN, Monte Carlo). Up to here, differential albedos are used for single reflections from walls in the NARCISSE-3 propagation albedo code developed at CEA and used for project calculations. For taking into account the neutron multiple reflections on lacunar medium walls, double energy-angle differential albedos are needed. TRIPOLI-4 neutral particle transport Monte Carlo code in three dimensional geometries, has been chosen to implement a double differential albedo calculus routine and therefore to generate albedo data for different kinds of medium. The surfacic estimator, which could be used, is not enough efficient because all neutrons do not contribute to the result. A new estimator is carried out. At each collision site, during the neutron history simulation, it allows to compute the probability of the neutron to go through the medium and to come through the reflection surface in the direction and at the energy considered. This estimator is about hundred times more efficient than the surfacic estimator. (author)
Calculation of fast neutron flux in reactor pressure tubes and experimental facilities
Energy Technology Data Exchange (ETDEWEB)
Barnett, P. C. [Canadian General Electric (Canada)
1968-07-15
The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.
A library of neutron data for calculating group constants
International Nuclear Information System (INIS)
Koshcheev, V.N.; Nikolaev, M.N.
1987-01-01
This paper describes the first version of a computerized library evaluated neutron data files (FOND) which includes data on the 67 most important nuclear reactor and radiation shielding materials. The data are represented in the ENDF/B format. The sources of data were the results of evaluations of data from differential neutron physics experiments conducted both in the USSR and abroad. The first version of the FOND library is not intended for use in reactor and shielding design calculations, but rather to serve as the basis for developing a corrected version which will guarantee adequate description of the results of a representative set of macroscopic experiments, and for generating multigroup constant systems in methodological research. (author)
Calculations of long-lived isomer production in neutron reactions
International Nuclear Information System (INIS)
Chadwick, M.B.; Young, P.G.
1992-01-01
We present theoretical calculations for the production of the long-lived isomers 93m Nb (1/2-, 16 yr), 121m Sn (11/2-, 55 yr), 166m Ho (7-, 1200 yr), 184m Re (8+, 165 d), 186m Re (8+, 2x10 5 yr), 178 Hf (16+, 31 yr), 179m Hf (25/2-, 25 d), and 192m Ir (9+, 241 yr), all of which pose potential radiation activation problems in nuclear fusion reactors if produced in 14-MeV neutron-induced reactions. We consider (n,2n), (n,n'), and (n,γ) production modes and compare our results both with experimental data (where available) and systematics. We also investigate the dependence of the isomeric cross section ratio on incident neutron energy for the isomers under consideration. The statistical Hauser-Feshbach plus preequilibrium code GNASH was used for the calculations. Where discrete state experimental information was lacking, rotational band members above the isomeric state, which can be justified theoretically but have not been experimentally resolved, were reconstructed. (author). 16 refs, 10 figs, 4 tabs
Calculation of neutron fluence-to-dose conversion factors for extremities
International Nuclear Information System (INIS)
Stewart, R.D.; Harty, R.; McDonald, J.C.; Tanner, J.E.
1993-04-01
The Pacific Northwest Laboratory is developing a standard for the performance testing of personnel extremity dosimeters for the US Department of Energy. Part of this effort requires the calculation of neutron fluence-to-dose conversion factors for finger and wrist extremities. This study focuses on conversion factors for two types of extremity models: namely the polymethyl methacrylate (PMMA) phantom (as specified in the draft standard for performance testing of extremity dosimeters) and more realistic extremity models composed of tissue-and-bone. Calculations for each type of model are based on both bare and D 2 O-moderated 252 Cf sources. The results are then tabulated and compared with whole-body conversion factors. More appropriate energy-averaged quality factors for the extremity models have also been computed from the neutron fluence in 50 equally spaced energy bins with energies from 2.53 x 10 -8 to 15 MeV. Tabulated results show that conversion factors for both types of extremity phantom are 15 to 30% lower than the corresponcung whole-body phantom conversion factors for 252 Cf neutron sources. This difference in extremity and whole-body conversion factors is attributable to the proportionally smaller amount of back-scattering that occurs in the extremity phantoms compared with whole-body phantoms
Calculations of mass and moment of inertia for neutron stars
International Nuclear Information System (INIS)
Moelnvik, T.; Oestgaard, E.
1985-01-01
Masses and moments of inertia for slowly-rotating neutron stars are calculated from the Tolman-Oppenheimer-Volkoff equations and various equations of state for neutron-star matter. We have also obtained pressure and density as a function of the distance from the centre of the star. Generally, two different equations of state are applied for particle densities n>0.47 fm -3 and n -3 . The maximum mass is, in our calculations for all equations of state except for the unrealistic non-relativistic ideal Fermi gas, given by 1.50 Msub(sun) 44 gxcm 2 45 gxcm 2 , which also seem to agree very well with 'experimental results'. The radius of the star corresponding to maximum mass and maximum moment of inertia is given by 8.2 km< R<10.0 km, but a smaller central density rhosub(c) will give a larger radius. (orig.)
Extensive Air Showers Detected by Aragats Neutron Monitor
International Nuclear Information System (INIS)
Badalyan, A.; Chilingarian, A.; Hovsepyan, G.; Grigoryan, A.; Khanikyants, Y.; Manukyan, A.; Pokhsraryan, D.; Soghomonyan, S.
2017-01-01
Extensive Air Shower (EAS) duration as registered by the surface particle detectors does not exceed a few tens of nanosecond. However, Neutron monitors containing plenty of absorbing matter can respond to EAS core traversal during 1 ∼ms by registering secondary slow neutrons born by EAS hadrons in the soil, walls of buildings and in the matter of detector itself. Thus, the time distribution of the pulses from the proportional counters of the neutron monitor after EAS propagation extends to ∼l ms, ∼5 orders of magnitude larger than the EAS passing time. The Aragats Neutron Monitor (ArNM) has a special option for the EAS core detection. In general, the dead time of NM is ∼1 ms that provides the one-to-one relation of incident hadrons and detector counts. The pulses generated by the neutrons possibly entering the proportional chamber after the first one will be neglected. In ArNM, we use several “electronic” dead times, and with the shortest one, 400 ns, the detector counts all pulses that enter the proportional chambers. If ArNM one-second time series corresponding to the shortest dead time contain much more signals (a neutron burst) than with l-ms dead time, then we conclude that the EAS core hits the detector. We assume that he distribution of registered burst multiplicities is proportional to the energy of the primary particle. The primary cosmic ray energy spectrum was obtained by the frequency analysis through the counting frequencies of the multiplicities of different magnitudes and relating them to the integral energy spectrum measured by the MAKET array at the same place several years ago. (author)
Calculation of neutron flux in the presence of a source
International Nuclear Information System (INIS)
Planchard, J.
1993-09-01
Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs
Air pollution measurements in a semi-arid zine, using neutron activation analysis
Energy Technology Data Exchange (ETDEWEB)
Shani, G; Cohen, D
1977-01-01
Neutron activation analysis was used for measurement of air pollution in a semi arid zone. After a careful selection of the air filter, air samples were filtered in the city of Beer-Sheva in the south of Israel. The sampling took place over a period of 2 months with each sampling period lasting 2 days. The samples were irradiated in a thermal neutron flux of the order of 10/sup 13/ n/cm/sup 2/ sec for 1 h and the gamma spectrum was measured several times. The concentration of elements was concluded from the measured gamma spectrum. Several conclusions were obtained by combining the weather conditions with the measurement results. The air pollution can be divided into several sources: dust of a certain origin containing Fe, Co, Cr, Sc, Th and Na (probably from the Dead Sea area), other dust sources containing Sb, Eu, and Hf, and urban pollution due to industry and transportation--Br, Hg. The air pollution has its maxima and minima according to weather or industrial conditions. The pollution requires 4 days to clear out. Other relations to weather conditions were also found.
Air pollution measurements in a semi-arid zone, using neutron activation analysis
Energy Technology Data Exchange (ETDEWEB)
Shani, G; Cohen, D [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering
1977-12-01
Neutron activation analysis was used for measurement of air pollution in a semi-arid zone. After a careful selection of the air filter, air samples were filtered in the city of Beer-Sheva in the south of Israel. The sampling took place over a period of 2 months with each sampling period lasting 2 days. The samples were irradiated in a thermal neutron flux of the order of 10/sup 13/ n/cm/sup 2/ sec for 1 h and the gamma spectrum was measured several times. The concentration of elements was concluded from the measured gamma spectrum. Several conclusions were obtained by combining the weather conditions with the measurement results. The air pollution can be divided into several sources: dust of a certain origin containing Fe, Co, Cr, Sc, Th and Na (probably from the Dead Sea area), other dust sources containing Sb, Eu, and Hf, and urban pollution due to industry and transportation-Br, Hg. The air pollution has its maxima and minima according to weather or industrial conditions. The pollution requires 4 days to clear out. Other relations to weather conditions were also found.
Air pollution measurements in a semi-arid zone, using neutron activation analysis
International Nuclear Information System (INIS)
Shani, G.; Cohen, D.
1977-01-01
Neutron activation analysis was used for measurement of air pollution in a semi-arid zone. After a careful selection of the air filter, air samples were filtered in the city of Beer-Sheva in the south of Israel. The sampling took place over a period of 2 months with each sampling period lasting 2 days. The samples were irradiated in a thermal neutron flux of the order of 10 13 n/cm 2 sec for 1 h and the gamma spectrum was measured several times. The concentration of elements was concluded from the measured gamma spectrum. Several conclusions were obtained by combining the weather conditions with the measurement results. The air pollution can be divided into several sources: dust of a certain origin containing Fe, Co, Cr, Sc, Th and Na (probably from the Dead Sea area), other dust sources containing Sb, Eu, and Hf, and urban pollution due to industry and transportation-Br, Hg. The air pollution has its maxima and minima according to weather or industrial conditions. The pollution requires 4 days to clear out. Other relations to weather conditions were also found. (Auth.)
International Nuclear Information System (INIS)
Rataj, J.
1993-10-01
The method of calculating neutron spectra and integral flux densities of neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor is outlined. The one-dimensional and two-dimensional calculations were performed by means of the DORT code in R, R-Z, and R-Θ geometries using the cross sections from the ELXSIR library. In the R-Θ geometry, the coupled neutron flux densities were determined. The calculated values of the maximum activation of detectors differ less than 15% from the values measured in surveillance specimens, which is within the limit of uncertainty associated with the position of the detector in the casing. The differences between the calculated and observed data behind the pressure vessel were below 4%. 10 tabs., 3 figs., 41 refs
Measured and calculated neutron yields for 100 MeV protons on thick targets of Pb and Li
International Nuclear Information System (INIS)
Jones, R.T.; Lone, M.A.; Okazaki, A.
1983-01-01
The neutron yield per proton from thick targets of lead and lithium irradiated with 100 MeV protons has been measured and calculated. The water bath method was used to measure the neutron production, and a Faraday cup for the beam current determination. Measured yields are 0.343 +- 0.021 for lead and 0.123 +- 0.007 for lithium. Corresponding yields calculated with the nucleon-meson transport code NMTC are 0.363 +- 0.002 and 0.160 +- 0.001. Measured and calculated thermal neutron distributions in the water bath are also compared
Research on GPU-accelerated algorithm in 3D finite difference neutron diffusion calculation method
International Nuclear Information System (INIS)
Xu Qi; Yu Ganglin; Wang Kan; Sun Jialong
2014-01-01
In this paper, the adaptability of the neutron diffusion numerical algorithm on GPUs was studied, and a GPU-accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. The IAEA 3D PWR benchmark problem was calculated in the numerical test. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. (authors)
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)
Comparison of Monte Carlo method and deterministic method for neutron transport calculation
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki
1987-01-01
The report outlines major features of the Monte Carlo method by citing various applications of the method and techniques used for Monte Carlo codes. Major areas of its application include analysis of measurements on fast critical assemblies, nuclear fusion reactor neutronics analysis, criticality safety analysis, evaluation by VIM code, and calculation for shielding. Major techniques used for Monte Carlo codes include the random walk method, geometric expression method (combinatorial geometry, 1, 2, 4-th degree surface and lattice geometry), nuclear data expression, evaluation method (track length, collision, analog (absorption), surface crossing, point), and dispersion reduction (Russian roulette, splitting, exponential transform, importance sampling, corrected sampling). Major features of the Monte Carlo method are as follows: 1) neutron source distribution and systems of complex geometry can be simulated accurately, 2) physical quantities such as neutron flux in a place, on a surface or at a point can be evaluated, and 3) calculation requires less time. (Nogami, K.)
International Nuclear Information System (INIS)
Avdic, S.; Pesic, M.
1992-01-01
The ORTEC 580 Neutron Spectrometer system contains a detector unit in diode coincidence arrangement for measurement of fast neutron spectrum in the energy range from 1 MeV to 14 MeV. Numerical code HE3 for computation of semiconductor 3 He detector efficiency in a collimated neutron beam is based on analytical method in infinite diode approximation and Monte Carlo method for real spectrometer geometry. Calculations are performed in the first collision approximation in the detector active volume including evaluation of correction factors. Accuracy of relative detector efficiency calculation is improved by using neutron cross section from nuclear library ENDF/B-6. (author)
International Nuclear Information System (INIS)
Spyrou, N.M.; Awotwi-Pratt, J.B.; Williams, A.M.
2004-01-01
An activation analysis facility based on an isotopic neutron source (185 GBq 241 Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, it was attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. (author)
Determination of mercury in air by adsorption on Hopcalite and by neutron activation analysis
International Nuclear Information System (INIS)
Leyni-Barbaz, D.; Zikovsky, L.; Poissant, L.
2002-01-01
A new method for the determination of mercury in air has been developed. It combines the adsorption of mercury on Hopcalite (a material approved for this purpose by the National Institute of Health of the United States) and its quantification by neutron activation. The concentrations of mercury in office air in Montreal, Canada, were determined by instrumental semiabsolute neutron activation analysis. They varied from 39 to 48 ng/m 3 . The results were compared with the concentrations of mercury in office air determined simultaneously at the same place by cold vapour atomic fluorescence spectrophotometry. A close correlation between the results of the 2 methods was obtained. The detection limit of our method is about 14 ng/m 3 . (author)
International Nuclear Information System (INIS)
Sakamoto, Yukio
2001-01-01
The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)
International Nuclear Information System (INIS)
Kodeli, I.; Diop, C.M.; Nimal, J.C.
1994-01-01
In the framework of future Boron Neutron Capture Therapy (BNCT) experiments, where cells and animals irradiations are planned at the research reactor of Strasbourg University, the feasibility to obtain a suitable epithermal neutron beam is investigated. The neutron fluence and spectra calculations in the reactor are performed using the 3D Monte Carlo code TRIPOLI-3 and the 2D SN code TWODANT. The preliminary analysis of Al 2 O 3 and Al-Al 2 O 3 filters configurations are carried out in an attempt to optimize the flux characteristics in the beam tube facility. 7 figs., 7 refs
Air ingression calculations for selected plant transients using MELCOR
International Nuclear Information System (INIS)
Kmetyk, L.N.
1994-01-01
Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression
Energy Technology Data Exchange (ETDEWEB)
Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R., E-mail: rosanemribeiro@oi.com.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2014-07-01
Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H{sub p} (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm{sup 3}, composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm{sup 2}). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)
International Nuclear Information System (INIS)
Schuerrer, F.
1980-01-01
For characterizing heterogene configurations of pebble-bed reactors the fine structure of the flux distribution as well as the determination of the macroscopic neutronphysical quantities are of interest. When calculating system parameters of Wigner-Seitz-cells the usual codes for neutron spectra calculation always neglect the modulation of the neutron flux by the influence of neighbouring spheres. To judge the error arising from that procedure it is necessary to determinate the flux distribution in the surrounding of a spherical fuel element. In the present paper an approximation method to calculate the flux distribution in the two-sphere model is developed. This method is based on the exactly solvable problem of the flux determination of a point source of neutrons in an infinite medium, which contains a spherical perturbation zone eccentric to the point source. An iteration method allows by superposing secondary fields and alternately satisfying the conditions of continuity on the surface of each of the two fuel elements to advance to continually improving approximations. (orig.) 891 RW/orig. 892 CKA [de
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
International Nuclear Information System (INIS)
Etemad, M.A.
1981-04-01
The one dimensional discrete ordinates code ANISN-F was used to calculate the thermal neutron flux distribution in water from a Ra-Be neutron source. The calculations were performed in order to investigate the different possibilities of the code as well as to verify the results of the calculations in terms of comparisons to corresponding experimental data. Two different group cross section libraries were used in the calculations and conclusions were drawn on the adequacy of these libraries for a fixed source type calculation. Furthermore, critically calculations were performed for an infinite homogeneous slab of multiplying material using different angular and spatial approximations. The results of these calculations were then compared to the corresponding results previously obtained at this department by a different method and a different code. (author)
International Nuclear Information System (INIS)
Dorlet, J.
1991-01-01
A describing algorithm furnishes the probabilities of having exactly N escaping neutrons in the descent of one fission neutron, using the punctual reactor model. Calculations can be performed even for N-values greater than 1000. Numerical results show that discrete neutrons counting is unfit to obtain the N mean value, even for very far subcritical devices. That mean value is still very used in the existing theoretical studies, because its obvious correlation with the device effective multiplication coefficient. For that reason modelling coincidence neutrons counting is not suitable using statistical moments approach, but only using probabilities it selves [fr
International Nuclear Information System (INIS)
Gritzay, O.; Kalchenko, O.; Klimova, N.; Razbudey, V.; Sanzhur, A.
2006-01-01
Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations
Model-Independent Calculation of Radiative Neutron Capture on Lithium-7
Rupak, Gautam; Higa, Renato
2011-01-01
The radiative neutron capture on lithium-7 is calculated model independently using a low-energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix elements. It depends on the poorly known p-wave effective range parameter
Calculation of spherical models of lead with a source of 14 MeV-neutrons
International Nuclear Information System (INIS)
Markovskij, D.V.; Borisov, A.A.
1989-01-01
Neutron transport calculations for spherical models of lead have been done with the one-dimensional code BLANK realizing the direct Monte Carlo method in the whole range of neutron energies and they are compared with the experimental results. 6 refs, 10 figs, 3 tabs
Neutrons characterization of the nuclear reactor Ian-R1 of Colombia
International Nuclear Information System (INIS)
Gonzalez P, L. X.; Martinez O, S. A.; Vega C, H. R.
2014-08-01
By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)
Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers
International Nuclear Information System (INIS)
Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.
1989-01-01
The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain
A sensitivity study on neutron flux variation due to 10B concentration in dose calculation for BNCT
International Nuclear Information System (INIS)
Jung, Sang Hoon
2006-02-01
The effects of inclusion of 10 B concentration on neutron flux and dose in dose calculation were studied. In order to provide the quantitative effects of inclusion of 10 B concentrations on depressions of neutron and photon flux and dose, the fluxes and doses with voxel head phantoms for various 10 B concentrations homogeneously distributed were calculated by using MCNPX simulations. A lithium target system and beam shaping assembly, which have been developed at the Hanyang University, were used as epithermal neutron beam. The calculation results show that the neutron flux at the center of the head phantom decreases by approximately 5.4% per 10 ppm of 10 B concentration in comparison with the neutron flux in the case of boron-free. It was also observed that the tissue dose at the center of the head phantom is depressed by approximately 4.7% per 10 ppm of the 10 B concentration and the tumor dose by approximately 5.3% per 10 ppm. According to depth of tumors, it was observed that the depressions of the doses in the tumors are ranged in 3.7 ∼ 9.2%. The dose calculations in the case of boron-free show that it is overestimated in comparison with the dose calculations in the cases of the inclusion of 10 B concentrations for the normal tissue and the tumors. Therefore, in dose calculation for BNCT, the depressions of neutron flux and dose should be considered. The results in this study are available to setting up the depression ratios which can be used for converting neutron and gamma fluxes and doses in phantom with boron free into the fluxes and doses in phantom with inclusion of 10 B concentrations in treatment. It is expected that the depression ratios is practicable to dose evaluation for BNCT
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
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)
International Nuclear Information System (INIS)
Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong
2001-01-01
Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf
MONTE CARLO CALCULATION OF THE ENERGY RESPONSE OF THE NARF HURST-TYPE FAST- NEUTRON DOSIMETER
Energy Technology Data Exchange (ETDEWEB)
De Vries, T. W.
1963-06-15
The response function for the fast-neutron dosimeter was calculated by the Monte Carlo technique (Code K-52) and compared with a calculation based on the Bragg-Gray principle. The energy deposition spectra so obtained show that the response spectra become softer with increased incident neutron energy ahove 3 Mev. The K-52 calculated total res nu onse is more nearly constant with energy than the BraggGray response. The former increases 70 percent from 1 Mev to 14 Mev while the latter increases 135 percent over this energy range. (auth)
Study on calculation methods for the effective delayed neutron fraction
International Nuclear Information System (INIS)
Irwanto, Dwi; Obara, Toru; Chiba, Go; Nagaya, Yasunobu
2011-03-01
The effective delayed neutron fraction β eff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate β eff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for β eff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate β eff with various definitions such as the fundamental value β 0 , the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of β eff , Nauchi's β eff and Meulekamp's β eff is approximately 10%. The fundamental value β 0 is quite larger than the others in several cases. For all the cases, Meulekamp's β eff is always higher than Nauchi's β eff . This is because Nauchi's β eff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's β eff does not include this parameter. Furthermore, we evaluate the multi-generation effect on β eff values and demonstrate that this effect should be considered to obtain the standard definition values of β eff . (author)
The calculated neutron response of a sphere with the multi-counters
International Nuclear Information System (INIS)
Li Taosheng; Yang Lianzhen; Li Dongyu
2004-01-01
Based on the difference of the neutron distribution in the moderator, three position sensitive proportional counters which are perpendicular to each other are inserted into the moderator. The energy responses with six spherical moderators and six incidence directions have been calculated by MCNP4A code. The calculated results for two divided region methods in the radial of the spherical moderator have been analyzed and compared. (authors)
Energy Technology Data Exchange (ETDEWEB)
Benomar, M
1998-09-01
The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.) 7 refs.
International Nuclear Information System (INIS)
Ainsworth, T.L.
1983-01-01
The Δ(1232) plays an important role in determining the properties of nuclear and neutron matter. The effects of the Δ resonance are incorporated explicitly by using a coupled channel formalism. A method for constraining a lowest order variational calculation, appropriate when nucleon internal degrees of freedom are made explicity, is presented. Different N-N potentials were calculated and fit to phase shift data and deuteron properties. The potentials were constructed to test the relative importance of the Δ resonance on nuclear properties. The symmetry energy and incompressibility of nuclear matter are generally reproduced by this calculation. Neutron matter results lead to appealing neutron star models. Fermi liquid parameters for 3 He are calculated with a model that includes both direct and induced terms. A convenient form of the direct interaction is obtained in terms of the parameters. The form of the direct interaction ensures that the forward scattering sum rule (Pauli principle) is obeyed. The parameters are adjusted to fit the experimentally determined F 0 /sup s/, F 0 /sup a/, and F 1 /sup s/ Landau parameters. Higher order Landau parameters are calculated by the self-consistent solution of the equations; comparison to experiment is good. The model also leads to a preferred value for the effective mass of 3 He. Of the three parameters only one shows any dependence on pressure. An exact sum rule is derived relating this parameter to a specific summation of Landau parameters
International Nuclear Information System (INIS)
Mueller, R.G.
1987-06-01
Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de
International Nuclear Information System (INIS)
Li Chunjuan; Liu Yi'na; Zhang Weihua; Wang Zhiqiang
2014-01-01
The manganese bath method for measuring the neutron emission rate of radionuclide sources requires corrections to be made for emitted neutrons which are not captured by manganese nuclei. The Monte Carlo particle transport code MCNP was used to simulate the manganese bath system of the standards for the measurement of neutron source intensity. The correction factors were calculated and the reliability of the model was demonstrated through the key comparison for the radionuclide neutron source emission rate measurements organized by BIPM. The uncertainties in the calculated values were evaluated by considering the sensitivities to the solution density, the density of the radioactive material, the positioning of the source, the radius of the bath, and the interaction cross-sections. A new method for the evaluation of the uncertainties in Monte Carlo calculation was given. (authors)
Calculation of the pulsed Feynman- and Rossi-alpha formulae with delayed neutrons
International Nuclear Information System (INIS)
Kitamura, Y.; Pazsit, I.; Wright, J.; Yamamoto, A.; Yamane, Y.
2005-01-01
In previous works, the authors have developed an effective solution technique for calculating the pulsed Feynman- and Rossi-alpha formulae. Through derivation of these formulae, it was shown that the technique can easily handle various pulse shapes of the pulsed neutron source. Furthermore, it was also shown that both the deterministic (i.e., synchronizing with the pulsing of neutron source) and stochastic (non-synchronizing) Feynman-alpha formulae can be obtained with this solution technique. However, for mathematical simplicity and the sake of insight, the formal derivation was performed in a model without delayed neutrons. In this paper, to demonstrate the robustness of the technique, the pulsed Feynman- and Rossi-alpha formulae were re-derived by taking one group of delayed neutrons into account. The results show that the advantages of this technique are retained even by inclusion of the delayed neutrons. Compact explicit formulae are derived for the Feynman- and Rossi-alpha methods for various pulse shapes and pulsing methods
Energy Technology Data Exchange (ETDEWEB)
Goldman, D T; Federighi, F D [Knolls Atomic Power Laboratory, General Electric Company, Schenectady, NY (United States)
1963-01-15
A model for the calculation of the scattering of thermal neutrons from chemical system was proposed by Nelkin. This model considered the actual dynamics of the scattering system as composed of a set of oscillatory motions, each describable by a Hamiltonian which commuted with each of the others. It was then possible to express the differential scattering cross-section in closed form. This model has been used to calculate the scattering of neutrons by water. Some care must be taken in performing the numerical integration over angle and energy. The scattering model has been extended to the calculation of neutron scattering from polyethylene C{sub n}H{sub 2n}. Analogous levels of polyethylene can be noted at 0.089 eV, 0.182 eV, 0.354 eV, and 0.533 eV. The differential and total cross-sections have been calculated for the scattering and the latter has been seen to be in reasonable agreement with experiment at room temperature. Scattering kernels have been calculated for a number of temperatures and where possible the results have been compared with experiment. In addition, neutron flux spectra and diffusion lengths have been calculated using the equations of reactor physics. Comparison of these Results with experimental data indicates that such integral measurements are indicative of at least the gross features of the scattering system and should be analysed in conduction with the detailed differential cross-section results. (author) [French] Nelkin a propose un modele pour calculer la diffusion de neutrons thermiques dans des systemes chimiques. Dans ce mod and le on considere que la dynamique reelle du systeme de diffusion se compose d'un ensemble de mouvements oscillatoires, chaque mouvement pouvant 6tre decrit par un hamiltonien commutant avec chacun des autres. Il est alors possible d'exprimer la section efficace differentielle de diffusion sous une forme fermee. Les auteurs ont employe ce modele pour calculer la diffusion des neutrons par l'eau. Il faut prendre
The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method
International Nuclear Information System (INIS)
Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang
2011-01-01
Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)
Activation measurements for fast neutrons. Part A. Sulfur (32P) activation
International Nuclear Information System (INIS)
Kerr, George D.
2005-01-01
The sulfur activation measurements at Hiroshima have been useful in refining and verifying the neutron source-term calculations for the Hiroshima bomb (Kerr and Pace 1988). For example, the Riken measurements of sulfur activation at Hiroshima were used to demonstrate that a two dimensional calculation of the neutron leakage was required for the Hiroshima bomb in order to accurately account for the blind spot in the neutron leakage through the nose of the device (Kerr 1982a,b; Kerr et al. 1983; Pace and Kerr 1984). The sulfur activation measurements at Hiroshima are discussed and compared in this section with the results of DS86 and DS02 calculations for the air-over-ground transport of neutrons at Hiroshima. (J.P.N.)
International Nuclear Information System (INIS)
Johnson, J.O.; Miller, L.F.; Kam, F.B.K.
1981-05-01
A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58 Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O.; Miller, L.F.; Kam, F.B.K.
1981-05-01
A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.
Microdosimetry of 14.7 MeV neutrons in tissue equivalent phantom
International Nuclear Information System (INIS)
Amols, H.I.
1974-01-01
An experimental and theoretical investigation has been made of energy deposition in tissue by neutrons. A one-half inch diameter Rossi type proportional counter was used to simulate a one-micron sphere of tissue. Event-size spectra were taken in air, and at various positions in a large volume of tissue equivalent fluid. From the raw spectra, LET distributions were determined, as well as dose fractions for protons, alphas, and heavy ions, and dose average and track-average LET values. The shape of the D(L) vs. LET curve is found to undergo significant change in the phantom due to moderation of the neutron beam. In addition, previous calculations of LET spectra in air are shown to be in error, and theoretical RBE and OER values, based on data from this experiment are in better agreement with biological results. A two-step theoretical calculation has also been carried out. An original Monte Carlo computer code was used to calculate neutron fluences in phantom (1), which were converted to LET distributions via standard algorithms (2). Agreement with experiment is very good, both in air and in phantom. Edge effects, backscatter effects, and effects of phantom size were also studied
International Nuclear Information System (INIS)
Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.
2010-01-01
Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are
FUP1--an unified program for calculating all fast neutron data of fissile nucleus
International Nuclear Information System (INIS)
Cai Chonghai; Zuo Yixin
1990-01-01
FUP1 is the first edition of an unified program for calculating all the fast neutron data in ENDF/B-4 format for fissile nucleus. Following data are calculated with FUP1 code: the total cross section, elastic scattering cross section, nonelastic cross section, total including up to 40 isolated levels and continuum state inelastic cross sections. In FUP1 the energy region of incident neutron is restricted to 10 Kev to 20 Mev. The advantages of this program are its perfect function, convenient to users and running very fast
Application of propagation calculations in air quality control
International Nuclear Information System (INIS)
Kuelske, S.
1993-01-01
This paper describes the development of a methodics of propagation calculation since the foundation of the 'Landesanstalt fuer Immissionsschutz' in 1963, and its practical use in air quality control. In this context, it deals with methods for calculating stack heights and the Gaussian propagation model adopted by the Technical Code on Clean Air, the propagation of flue gas emitted via cooling towers, the propagation of accidentally released substances, odours and automobile emissions, and with flow and propagation calculation for the proximity zone of buildings as well as for meteorological fields with space and time variations and topographically varied terrain. (orig.) [de
Descartes: a new generation system for neutronic calculations
International Nuclear Information System (INIS)
Calvin, Ch.
2005-01-01
Descartes is a common project between CEA, Framatome and EDF for the development of a new generation system for neutronic calculations. The main objectives which have leaded the design of the platform are the following: - flexible: from best-estimate calculations to industrial design; - open: easy coupling with other disciplines (thermo mechanics, thermal hydraulics); - enlarged scope: criticality, shielding, all types of reactors; - robust: well known behavior in its field of application; - safe: qualified and uncertainties assessment; and - User-friendly: user interface, databases; Descartes is based on the object oriented method using UML design and programmed in C++ and the Python interpreted script language. We will present in this paper the general architecture of the platform and the internal data model used which allows the definition of common exchange structures between solvers and the different modules which can be used either for lattice or core calculations. In a second time we will present a short description of the main solvers implemented within the Descartes platform. We will conclude with some first results of industrial PWR calculations. (author)
International Nuclear Information System (INIS)
Verschuur, K.A.
1996-10-01
Neutron-transport calculations with the FURNACE(2) program system, in support of the Neutron Diagnostic Group at JET, have been performed since 1980, i.e. since the construction phase of JET. FURNACE(2) is a ray-tracing/multiple-reflection transport program system for toroidal geometries, that orginally was developed for blanket neutronics studies and which then was improved and extended for application to the neutron-diagnostics at JET. (orig./WL)
Energy Technology Data Exchange (ETDEWEB)
Mooney, L. G.
1963-06-15
Calculations of neutron penetration probabilities were performed to evaluate the Monte Carlo Multilayer Slab Penetration Procedure. A 9-in. polyethylene alab was chosen for the calculations and results were compared with experimental data. The calculated and measured dose rates agree within 20% for all exit polar angles. The calculations indicate that incident neutrons with energies less than 2.5 Mev do not contribute significantly to the transmitted dose rate. (auth)
International Nuclear Information System (INIS)
Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji
1999-03-01
As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times
Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans
2006-02-01
GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.
Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis
International Nuclear Information System (INIS)
Duckic, P.; Matijevic, M.; Grgic, D.
2016-01-01
In this paper initial set of data for neutron buildup factors determination using Support Vector Regression (SVR) method is prepared. The performance of SVR technique strongly depends on the quality of information used for model training. Thus it is very important to provide representable data to the SVR. SVR is a supervised type of learning so it demands data in the input/output form. In the case of neutron buildup factors estimation, the input parameters are the incident neutron energy, shielding thickness and shielding material and the output parameter is the neutron buildup factor value. So far the initial sets of data for different shielding configurations have been obtained using SCALE4.4 sequence SAS3. However, this results were obtained using group constants, thus the incident neutron energy was determined as the average value for each energy group. Obtained this way, the data provided to the SVR are fewer and therefore insufficient. More valuable information is obtained using SCALE6.2beta5 sequence MAVRIC which can perform calculations for the explicit incident neutron energy, which leads to greater maneuvering possibilities when active learning measures are employed, and consequently improves the quality of the developed SVR model.(author).
Influence of the number of energy groups on the accuracy of neutron fluence calculations
International Nuclear Information System (INIS)
Barz, H.U.; Konheiser, J.
1999-01-01
The question how many groups are necessary to obtain all needed integral quantities for the neutron load of pressure vessels and detector positions outside the vessel with sufficient accuracy is of general interest. Until now, there are no systematic investigations on this question. In principle 3-dimensional consideration is required for such neutron load calculations. Therefore, an estimation of the needed number of groups can be of interest to minimize calculation time. One general problem is the P L -approximation of the angular distributions for the transfers between different groups. For elastic scattering this P L -approximation becomes poorer with increasing number of groups. As deterministic methods generally use the P L -approximation they cannot be used for investigations of the errors caused by the group approximation. We have investigated this problem applying group Monte-Carlo but nearly exact representation of this elastic slowing down without P L -approximation. The calculations were directed to assess the neutron fluence of a Russian WWER-1000 reactor. For that a simplified geometrical model of this reactor type has been used. (orig.)
International Nuclear Information System (INIS)
Bosevski, T.
1986-01-01
An improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell has been developed. Expanding the neutron flux and source into a series of even powers of the radius, one' gets a convenient method for integration of the one-energy group integral transport equation. It is shown that it is possible to perform an analytical integration in the x-y plane in one variable and to use the effective Gaussian integration over another one. Choosing a convenient distribution of space points in fuel and moderator the transport matrix calculation and cell reaction rate integration were condensed. On the basis of the proposed method, the computer program DISKRET for the ZUSE-Z 23 K computer has been written. The suitability of the proposed method for the calculation of the thermal-neutron-flux distribution in a reactor cell can be seen from the test results obtained. Compared with the other collision probability methods, the proposed treatment excels with a mathematical simplicity and a faster convergence. (author)
Energy Technology Data Exchange (ETDEWEB)
Kamykowski, E.A. (Grumman Corporate Research Center, Bethpage, NY (United States))
1992-07-15
Boron-loaded scintillators offer the potential for neutron spectrometers with a simplified, peak-shaped response. The Monte Carlo code, MCNP, has been used to calculate the detector characteristics of a scintillator made of a boron-loaded plastic, BC454, for neutrons between 1 and 7 MeV. Comparisons with measurements are made of spectral response for neutron energies between 4 and 6 MeV and of intrinsic efficiencies for neutrons up to 7 MeV. In order to compare the calculated spectra with measured data, enhancements to MCNP were introduced to generate tallies of light output spectra for recoil events terminating in a final capture by {sup 10}B. The comparison of measured and calculated spectra shows agreement in response shape, full width at half maximum, and recoil energy deposition. Intrinsic efficiencies measured to 7 MeV are also in agreement with the MCNP calculations. These results validate the code predictions and affirm the value of MCNP as a useful tool for development of sensor concepts based on boron-loaded plastics. (orig.).
Neutron physics calculation for WWER-1000 absorber element lifetime determination
International Nuclear Information System (INIS)
Kurakin, K.Yu.; Kushmanov, S.A.
2009-01-01
Absorber element with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 - 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development. The paper deals with the issues related to calculation determination of main factors that influence AE service life limitation - neutron flux and fluence onto absorbing and structural materials during extended service life. (Authors)
International Nuclear Information System (INIS)
Sekimoto, Hiroshi
1989-01-01
The tritium production density, kerma heat production density, dose and certain integral values of scalar neutron spectra in bare and graphite-reflected lithium-fluoride piles irradiated with D-T neutrons were evaluated from the pulse height distribution of a miniature NE213 neutron spectrometer with UFO data processing code, and compared with the values calculated with MORSE-CV Monte Carlo code. (author). 8 refs.; 1 fig.; 2 tabs
FLUKA Calculation of the Neutron Albedo Encountered at Low Earth Orbits
Claret, Arnaud; Combier, Natacha; Ferrari, Alfredo; Laurent, Philippe
2014-01-01
This paper presents Monte-Carlo simulations based on the Fluka code aiming to calculate the contribution of the neutron albedo at a given date and altitude above the Earth chosen by the user. The main input parameters of our model are the solar modulation affecting the spectra of cosmic rays, and the date of the Earth’s geomagnetic fi eld. The results consist in a two-parameter distribution, the neutron energy and the angle to the tangent plane of the sphere containing the orbi t of interest, and are provided by geographical position above the E arth at the chosen altitude. This model can be used to predict the te mporal variation of the neutron fl ux encountered along the orbit, and thus constrain the determination of the instrumental backg round noise of space experiments in low earth orbit.
Neutron shielding performance of water-extended polyester
International Nuclear Information System (INIS)
Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.
2006-01-01
A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)
Neutron shielding performance of water-extended polyester
Energy Technology Data Exchange (ETDEWEB)
Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)
2006-07-01
A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)
International Nuclear Information System (INIS)
Tel, E.; Kisoglu, H. F.; Topaksu, A. K.; Aydin, A.; Kaplan, A.
2007-01-01
There are several new technological application fields of fast neutrons such as accelerator-driven incineration/ transmutation of the long-lived radioactive nuclear wastes (in particular transuranium nuclides) to short-lived or stable isotopes by secondary spallation neutrons produced by high-intensity, intermediate-energy, charged-particle beams, prolonged planetary space missions, shielding for particle accelerators. Especially, accelerator driven subcritical systems (ADS) can be used for fission energy production and /or nuclear waste transmutation as well as in the intermediate-energy accelerator driven neutron sources, ions and neutrons with energies beyond 20 MeV, the upper limit of exiting data files that produced for fusion and fission applications. In these systems, the neutron scattering cross sections and emission differential data are very important for reactor neutronics calculations. The transition rate calculation involves the introduction of the parameter of mean free path determines the mean free path of the nucleon in the nuclear matter. This parameter allows an increase in mean free path, with simulation of effect, which is not considered in the calculations, such as conservation of parity and angular momentum in intra nuclear transitions. In this study, we have investigated the multiple preequilibrium matrix element constant from internal transition for Uranium, Thorium, (n,xn) neutron emission spectra. The neutron-emission spectra produced by (n,xn) reactions on nuclei of some target (for spallation) have been calculated. In the calculations, we have used the geometry dependent hybrid model and the cascade exciton model including the effects of the preequilibrium. The pre-equilibrium direct effects have been examined by using full exciton model. All calculated results have been compared with the experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other
Improvements in the model of neutron calculations for research reactors
International Nuclear Information System (INIS)
Calzetta, Osvaldo; Leszczynski, Francisco
1987-01-01
Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author) [es
Calculation of the Chilling Requirement for Air Conditioning in the Excavation Roadway
Directory of Open Access Journals (Sweden)
Yueping Qin
2015-10-01
Full Text Available To effectively improve the climate conditions of the excavation roadway in coal mine, the calculation of the chilling requirement taking air conditioning measures is extremely necessary. The temperature field of the surrounding rock with moving boundary in the excavation roadway was numerically simulated by using finite volume method. The unstable heat transfer coefficient between the surrounding rock and air flow was obtained via the previous calculation. According to the coupling effects of the air flow inside and outside air duct, the differential calculation mathematical model of air flow temperature in the excavation roadway was established. The chilling requirement was calculated with the selfdeveloped computer program for forecasting the required cooling capacity of the excavation roadway. A good air conditioning effect had been observed after applying the calculated results to field trial, which indicated that the prediction method and calculation procedure were reliable.
Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film
International Nuclear Information System (INIS)
Kalsi, P.C.
2000-01-01
The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)
The background cross section method for calculating the epithermal neutron spectra
International Nuclear Information System (INIS)
Martinez, A.S.
1983-01-01
We have developed a new methodology to the multigroup constants calculations, for thermal and fast reactors. The method to obtain the constants is extremely fast and simple, and it avoid repeated computations of the detailed neutron spectrum for different cell configurations (composition, geometry and temperature). (author) [pt
Cronos 2: a neutronic simulation software for reactor core calculations
International Nuclear Information System (INIS)
Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.
1999-01-01
The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)
Energy Technology Data Exchange (ETDEWEB)
Mansy, Muhammad S., E-mail: mmansy88@asrt.sci.eg [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Radioactive Waste Management Unit, Hot Labs Centre, Atomic Energy Authority, Cairo (Egypt); Adib, M.; Habib, N. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Bashter, I.I. [Physics Department, Faculty of Science, Zagazig University (Egypt); Morcos, H.N.; El-Mesiry, M.S. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt)
2016-10-01
Highlights: • Slow neutron cross-section calculation for poly- and mono-crystalline materials. • Monochromatic features of PbF{sub 2} and NaCl mono-crystals. • Characterization of poly- and mono-crystal filters used in neutron diffraction. • Computer code developed calculates neutron cross-section, transmission & reflection. - Abstract: A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF{sub 2} poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF{sub 2} cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF{sub 2} (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.
Theoretical calculation of solid particles deposition from the air
Directory of Open Access Journals (Sweden)
Bobro Milan
2002-03-01
Full Text Available This paper presents the calculation of harmful substance deposition (air pollution from the point source (Slanèo, et al., 2001 using equation (1. The point source shall be understood as e.g. chimneys of factory, heat plant, incinerator, boiler plant, local heating plant, etc.The theoretical calculation of concentration (1, or deposition (8 is based on the study of transfer and dispersion of pollution in air (Slanèo, et al., 2000a. The movement of pollution in air consists of a movement of the air itself and a relative movement of pollution particles and air, while the movement of harmful substance in the smoke trail is under the influence of turbulent diffusion, convection and gravitation. Molecular diffusion is not important in this process. When calculating concentrations (1 and deposition (8 of air pollution on a particular place near the source, it is assumed that the air speed is constant, the direction of wind does not change with the height and the source of air pollution is time-constant. The change in the wind speed with the height depends on the stability class of atmosphere (temperature gradient (Slanèo, et al., 2000a and it is calculated using equation (10.The theoretical calculation of concentration and or deposition of harmful substance from the point source (1 and (8 shall be applied if the harmful substance particles, which leave the source, have the same density (composition, shape (spherical and size.The experimental observations of dust deposition showed the significance of 0.1-20 µm particles. The application of equation (1 to calculate the concentration is conditioned, in addition to the recognition of source parameters and meteorological conditions, by the recognition of the particle sedimentation speed, which changes with the size of particle radius (2.For a practical calculation of deposition it is therefore necessary to know the differential distribution function f(r of particle radii, which can be made on the basis
International Nuclear Information System (INIS)
Tayloe, R.W. Jr.; Brown, A.S.; Dobelbower, M.C.; Woollard, J.E.
1997-03-01
Neutron-sensitive radiation detectors are used in the Portsmouth Gaseous Diffusion Plant's (PORTS) criticality accident alarm system (CAAS). The CAAS is composed of numerous detectors, electronics, and logic units. It uses a telemetry system to sound building evacuation horns and to provide remote alarm status in a central control facility. The ANSI Standard for a CAAS uses a free-in-air dose rate to define the detection criteria for a minimum accident-of-concern. Previously, the free-in-air absorbed dose rate from neutrons was used for determining the areal coverge of criticality detection within PORTS buildings handling fissile materials. However, the free-in-air dose rate does not accurately reflect the response of the neutron detectors in use at PORTS. Because the cost of placing additional CAAS detectors in areas of questionable coverage (based on a free-in-air absorbed dose rate) is high, the actual response function for the CAAS neutron detectors was determined. This report, which is organized into three major sections, discusses how the actual response function for the PORTS CAAS neutron detectors was determined. The CAAS neutron detectors are described in Section 2. The model of the detector system developed to facilitate calculation of the response function is discussed in Section 3. The results of the calculations, including confirmatory measurements with neutron sources, are given in Section 4
Uncertainty evaluatins of CASMO-3/MASTER system for PWR core neutronics calculations
International Nuclear Information System (INIS)
Song, Jae Seung; Kim, Kang Seog; Lee, Kibog; Park, Jin Ha; Zee, Sung Quun
1996-01-01
Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth
Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor
International Nuclear Information System (INIS)
Czernecki, Sebastien
1998-01-01
The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr
International Nuclear Information System (INIS)
Shen Chunxia; Qian Jianfu; Zhang Wenzhong
2003-01-01
This paper mainly calculate neutron activation discriminating the chemical weapons underground, and analyses the factors that soil influence discrimination, finally we conclude soil can not influence discrimination. (authors)
Air and smear sample calculational tool for Fluor Hanford Radiological control
International Nuclear Information System (INIS)
BAUMANN, B.L.
2003-01-01
A spreadsheet calculation tool was developed to automate the calculations performed for determining the concentration of airborne radioactivity and smear counting as outlined in HNF--13536, Section 5.2.7, ''Analyzing Air and Smear Samples''. This document reports on the design and testing of the calculation tool. Radiological Control Technicians (RCTs) will save time and reduce hand written and calculation errors by using an electronic form for documenting and calculating work place air samples. Current expectations are RCTs will perform an air sample and collect the filter or perform a smear for surface contamination. RCTs will then survey the filter for gross alpha and beta/gamma radioactivity and with the gross counts utilize either hand calculation method or a calculator to determine activity on the filter. The electronic form will allow the RCT with a few key strokes to document the individual's name, payroll, gross counts, instrument identifiers; produce an error free record. This productivity gain is realized by the enhanced ability to perform mathematical calculations electronically (reducing errors) and at the same time, documenting the air sample
Theoretical calculation and evaluation of complete neutron data for natural niobium
International Nuclear Information System (INIS)
Ma Gonggui; Zou Yiming; Wang Shiming
1990-07-01
An evaluation of a complete neutron nuclear data for natural niobium has been finished on the data measured by experiments up to 1989 and theoretical calculations with program MUP2 and AUJP. The purpose of present work is to build CENDL-2 databank (Chinese Evaluation Nuclear Data Library, second version) which replaces the CENDL-1 (first version of CENDL). The neutron energy for niobium is in the range of 10 -5 eV to 20 MeV. Data of cross section include total, elastic, nonelastic, total elastic, inelastic cross section to 13 discrete levels, inelastic continuum, (n,2n), (n,3n), (n,n ' α) + (n,αn ' ), (n,n ' p) + (n,pn ' ), (n,n ' d) + (n,dn ' ), (n,p), (n,d), (n,t), (n,α) and capture cross sections. Data for MT 251,252 and 253 as well as angular distributions and energy spectra of secondary neutrons are also given
232Th and 238U neutron emission cross section calculations and analysis of experimental data
International Nuclear Information System (INIS)
Tel, E.
2004-01-01
In this study, pre-equilibrium neutron-emission spectra produced by (n,xn) reactions on nuclei 2 32Th and 2 38U have been calculated. Angle-integrated cross sections in neutron induced reactions on targets 2 32Th and 2 38U have been calculated at the bombarding energies up to 18 MeV. We have investigated multiple pre-equilibrium matrix element constant from internal transition for 2 32Th (n,xn) neutron emission spectra. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach-Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n,xn) reaction on 232 T h and 2 38U, with the incident energy from 2 Me V to 18 Me V, the importance of multiple pre-equilibrium emission can be seen cleady. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other
Generalized diffusion theory for calculating the neutron transport scalar flux
International Nuclear Information System (INIS)
Alcouffe, R.E.
1975-01-01
A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)
NRSC, Neutron Resonance Spectrum Calculation System
International Nuclear Information System (INIS)
Leszczynski, Francisco
2004-01-01
1 - Description of program or function: The NRSC system is a package of four programs for calculating detailed neutron spectra and related quantities, for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening and unresolved resonance level. 2 - Methods: NRSC consists of four programs: GEXSCO, RMET21, ALAMBDA and WLUTIL. GEXSCO prepares the nuclear data from ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening or unresolved resonance level for RMET21 input. RMET21 calculates spectra and related quantities for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using slowing-down algorithms and, in the case of pin cells, the collision probability method. ALAMBDA obtains lambda factors (Goldstein-Cohen intermediate resonance factors in the formalism of WIMSD code) of different isotopes for including on WIMSD-type multigroup libraries for WIMSD or other cell-codes, from output of RMET21 program. WLUTIL is an auxiliary program for extracting tabulated parameters related with RMET21 program calculations from WIMSD libraries for comparisons, and for producing new WIMSD libraries with parameters calculated with RMET21 and ALAMBDA programs. 3 - Restrictions on the complexity of the problem: GEXSCO program has fixed array dimensions that are suitable for processing all reasonable outputs from nuclear data pre-processing programs. RMET21 program uses variable dimension method from a fixed general array. ALAMBDA and WLUTIL programs have fixed arrays that are adapted to standard WIMSD libraries. All programs can be easily modified to adapt to special requirements
Verification of RRC Ki code package for neutronic calculations of WWER core with GD
International Nuclear Information System (INIS)
Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.
2001-01-01
The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)
System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors
International Nuclear Information System (INIS)
Moiseyev, A.V.
2008-01-01
There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)
System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors
Energy Technology Data Exchange (ETDEWEB)
Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)
2008-07-01
There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)
Neutron shielding performance of water-extended polyester
International Nuclear Information System (INIS)
Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.
2006-01-01
A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)
Neutronic Design Calculations on Moderators for the Spallation Neutron Source (SNS)
International Nuclear Information System (INIS)
Murphy, D.B.
1999-01-01
The Spallation Neutron Source (SNS) to be built at the Oak Ridge National Laboratory will provide an intense source of neutrons for a large variety of experiments. It consists of a high-energy (1-GeV) and high-power (∼1-MW) proton accelerator, an accumulator ring, together with a target station and an experimental area. In the target itself, the proton beam will produce neutrons via the spallation process and these will be converted to low-energy ( 2 O moderators. Extensive engineering design work has been conducted on the moderator vessels. For our studies we have produced realistic neutronic representations of these moderators. We report on neutronic studies conducted on these representations of the moderators using Monte Carlo simulation techniques
International Nuclear Information System (INIS)
Bozoian, M.; Arthur, E.D.; Perry, R.T.; Wilson, W.B.; Young, P.G.
1988-01-01
Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab
International Nuclear Information System (INIS)
Pla, R.R.; Moreno, M.A.; Adler, M.
2000-01-01
Biomonitors were used as part of a pollution study of Buenos Aires city atmosphere under the International Atomic Energy Agency Research Contract ARG 7251, from the Co-ordinated Research Programme on Applied Research on Air Pollution using Nuclear Related Analytical Techniques. Lichens were primarily selected as indicators. Two different approaches were conducted, direct sampling of Parmotrema reticulatum, at a few places and the use of lichen bags, filled with Usnea sulcata from a northern national park, and hung at different sites. Simultaneously, tree bark was tried as biomonitor. Platanus acerifolia and Melia azedarach were selected as candidates, for being the most common trees in the city, but only P. acerifolia was analyzed. All the samples were analyzed using instrumental neutron activation analysis at the Ezeiza Atomic Centre of the National Atomic Energy Commission. RA-3) reactor was used for the irradiations, determining: As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, K, La, Lu, Na, Rb, Sb, Sc, Sm, Ta, Tb, Th, U, Yb and Zn. Concentration values for P. reticulatum compared well with values from literature. For U. sulcata differences were found among the tested sites and also, for some elements an increasing trend with time was observed. Enrichment factors calculated using Sc as reference and Mason's crustal average concentrations showed vehicules and refuse incineration as contributing sources to the aerosol. Tree bark from Buenos Aires and from a smaller city with mainly agricultural activities were analyzed and the results are coincident with those from lichens. This work is the first and preliminar contribution to the study of Buenos Aires aerosol using biomonitors. (author)
International Nuclear Information System (INIS)
Bondars, Kh.Ya.; Lapenas, A.A.
1981-01-01
We adapted or used on ES EhVM, operating under the control of OS ES, the currently most common algorithms for calculating neutron spectra from measured reaction rates. These programs, together with the neutron cross-section and spectrum libraries, are part of the computerized information system SAIPS. The present article descibes the basic mathematical concepts used in the algorithms of the SAIPS calculation programs
MURALB - a programme for calculating neutron fluxes in many groups
International Nuclear Information System (INIS)
MacDougall, J.
1977-09-01
The program MURALB solves the multi-group transport equation (with no upscatter) in many equal lethargy groups to produce neutron fluxes in these groups. The code has been made very flexible by confining the spatial flux solution to a single subroutine which takes as input the cross section data and source for a single group and calculates the flux for that group. In this way by supplying different versions of this routine different geometries and methods of solution of the transport equation may be treated. At present plane, cylindrical and spherical diffusion theory and collision probability solutions are available, together with a two region collision probability solution for a rod in a square cell. There is no basic restriction to one dimension but the practical size of problem tends to be limited to about 30 spatial regions by core storage requirements. In addition to the flux solution, the code calculates neutron balance, reaction rates and few groups cross sections for each mesh region, together with the values averaged over the system (cell or reactor). The program is available both as a stand-alone code and integrated into the COSMOS system. (author)
Verification of the DUCT-III for calculation of high energy neutron streaming
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Nakano, Hideo; Nakashima, Hiroshi; Sasamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tayama, Ryu-ichi; Handa, Hiroyuki; Hayashi, Katsumi [Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Shin, Kazuo [Kyoto Univ., Kyoto (Japan)
2003-03-01
A large number of radiation streaming calculations under a variety of conditions are required as a part of shielding design for a high energy proton accelerator facility. Since sophisticated methods are very time consuming, simplified methods are employed in many cases. For accuracy evaluation of a simplified code DUCT-III for high energy neutron streaming calculations, two kinds of benchmark problems based on the experiments were analyzed. Through comparison of the DUCT-III calculations with both the measurements and the sophisticated Monte Carlo calculations, DUCT-III was seen reliable enough for applying to the shielding design for the Intense Proton Accelerator Facility. (author)
Verification of the DUCT-III for calculation of high energy neutron streaming
Masukawa, F; Hayashi, K; Hirayama, H; Nakano, H; Nakashima, H; Sasamoto, N; Shin, K; Tayama, R I
2003-01-01
A large number of radiation streaming calculations under a variety of conditions are required as a part of shielding design for a high energy proton accelerator facility. Since sophisticated methods are very time consuming, simplified methods are employed in many cases. For accuracy evaluation of a simplified code DUCT-III for high energy neutron streaming calculations, two kinds of benchmark problems based on the experiments were analyzed. Through comparison of the DUCT-III calculations with both the measurements and the sophisticated Monte Carlo calculations, DUCT-III was seen reliable enough for applying to the shielding design for the Intense Proton Accelerator Facility.
Mansy, Muhammad S.; Adib, M.; Habib, N.; Bashter, I. I.; Morcos, H. N.; El-Mesiry, M. S.
2016-10-01
A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF2 poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF2 cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF2 (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.
The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity
International Nuclear Information System (INIS)
Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.
1987-04-01
A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry
International Nuclear Information System (INIS)
Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.
1987-04-01
A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable
International Nuclear Information System (INIS)
Taryo, Taswanda
2002-01-01
Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect
International Nuclear Information System (INIS)
Paredes G, L.C.; Balcazar G, M.; Francois L, J.L.; Azorin N, J.
2008-01-01
The results of the neutrons yield in different components of the bolster of an accelerator Varian Clinac 2100C of 18 MV for radiotherapy are presented, which contribute to the radiation of flight of neutrons in the patient and bolster planes. For the calculation of the neutrons yield, a simplified geometric model of spherical cell for the armor-plating of the bolster with Pb and W was used. Its were considered different materials for the Bremsstrahlung production and of neutrons produced through the photonuclear reactions and of electro disintegration, in function of the initial energy of the electron. The theoretical result of the total yield of neutrons is of 1.17x10 -3 n/e, considering to the choke in position of closed, in the patient plane with a distance source-surface of 100 cm; of which 15.73% corresponds to the target, 58.72% to the primary collimator, 4.53% to the levelled filter of Fe, 4.87% to the levelled filter of Ta and 16.15% to the closed choke. For an initial energy of the electrons of 18 MeV, a half energy of the neutrons of 2 MeV was obtained. The calculated values for radiation of experimental neutrons flight are inferior to the maxima limit specified in the NCRP-102 and IEC-60601-201.Ed.2.0 reports. The absorbed dose of neutrons determined through the measurements with TLD dosemeters in the isocenter to 100 cm of the target when the choke is closed one, is approximately 3 times greater that the calculated for armor-plating of W and 1.9 times greater than an armor-plating of Pb. (Author)
International Nuclear Information System (INIS)
Trotter, D.E. Gonzalez; Meneses, F. Salinas; Tornow, W.; Crowell, A.S.; Howell, C.R.; Schmidt, D.; Walter, R.L.
2009-01-01
The methods employed and the results obtained from measurements and calculations of the detection efficiency for the neutron detectors used at Triangle Universities Nuclear Laboratory (TUNL) in the simultaneous determination of the 1 S 0 neutron-neutron and neutron-proton scattering lengths a nn and a np , respectively, are described. Typical values for the detector efficiency were 0.3. Very good agreement between the different experimental methods and between data and calculation has been obtained in the neutron energy range below E n =13MeV.
International Nuclear Information System (INIS)
Reupke, W.A.; Davidson, J.N.; Muir, D.W.
1982-01-01
The authors present algorithms, describe a computer program, and gives a computational procedure for the statistical consistency analysis of neutron cross-section data, S /SUB N/ calculations, and measured tritium production in 14-MeV neutron-driven integral assemblies. Algorithms presented include a reduced matrix manipulation technique suitable for manygroup, 14-MeV neutron transport calculations. The computer program incorporates these algorithms and is expanded and improved to facilitate analysis of such integral experiments. Details of the computational procedure are given for a natural lithium deuteride experiment performed at the Los Alamos National Laboratory. Results are explained in terms of calculated cross-section sensitivities and uncertainty estimates. They include a downward adjustment of the 7 Li(n,xt) 14-MeV cross section from 328 + or - 22 to 284 + or - 24 mb, which is supported by the trend of recent differential and integral measurements. It is concluded that with appropriate refinements, the techniques of consistency analysis can be usefully applied to the analysis of 14-MeV neutron-driven tritium production integral experiments
An analytical method for neutron thermalization calculations in heterogenous reactors
Energy Technology Data Exchange (ETDEWEB)
Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)
1965-07-01
It is well known that the use of the diffusion approximation for stu
An analytical method for neutron thermalization calculations in heterogenous reactors
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1965-01-01
It is well known that the use of the diffusion approximation for studying neutron thermalization in heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations
Optimization of the neutron calculation model for the RA-6 reactor
International Nuclear Information System (INIS)
Coscia, G.A.
1981-01-01
A model for the neutronic calculation of the RA-6 reactor which includes the codes ANISN and EQUIPOSE is analyzed. Starting with a brief description of the reactor, the core and its parts, the general scheme of calculation applied is presented. The fuel elements used were those which are utilized in the RA-3 reactor; this is of the MTR type with 90% enriched uranium. With the approximations used, an analysis of such model of calculation was made, trying to optimize it by reducing, if possible, the calculation time without loosing accuracy. In order to improve the calculation model, it is recomended a cross section data library specific for the enrichment of the fuel considered 90% and the incorporation of a more advanced code than EQUIPOISE which would be DIXYBAR. (M.E.L.) [es
Neutron spectra calculation in material in order to compute irradiation damage
International Nuclear Information System (INIS)
Dupont, C.; Gonnord, J.; Le Dieu de Ville, A.; Nimal, J.C.; Totth, B.
1982-01-01
This short presentation will be on neutron spectra calculation methods in order to compute the damage rate formation in irradiated structure. Three computation schemes are used in the French C.E.A.: (1) 3-dimensional calculations using the line of sight attenuation method (MERCURE IV code), the removal cross section being obtained from an adjustment on a 1-dimensional transport calculation with the discrete ordinate code ANISN; (2) 2-dimensional calculations using the discrete ordinates method (DOT 3.5 code), 20 to 30 group library obtained by collapsing the 100 group a library on fluxes computed by ANISN; (3) 3-dimensional calculations using the Monte Carlo method (TRIPOLI system). The cross sections which originally came from UKNDL 73 and ENDF/B3 are now processed from ENDF B IV. (author)
A new method for calculation of an air quality index
Energy Technology Data Exchange (ETDEWEB)
Ilvessalo, P. [Finnish Meteorological Inst., Helsinki (Finland). Air Quality Dept.
1995-12-31
Air quality measurement programs in Finnish towns have expanded during the last few years. As a result of this it is more and more difficult to make use of all the measured concentration data. Citizens of Finnish towns are nowadays taking more of an interest in the air quality of their surroundings. The need to describe air quality in a simplified form has increased. Air quality indices permit the presentation of air quality data in such a way that prevailing conditions are more easily understandable than when using concentration data as such. Using an air quality index always means that some of the information about concentrations of contaminants in the air will be lost. How much information is possible to extract from a single index number depends on the calculation method. A new method for the calculation of an air quality index has been developed. This index always indicates the overstepping of an air quality guideline level. The calculation of this air quality index is performed using the concentrations of all the contaminants measured. The index gives information both about the prevailing air quality and also the short-term trend. It can also warn about the expected exceeding of guidelines due to one or several contaminants. The new index is especially suitable for the real-time monitoring and notification of air quality values. The behaviour of the index was studied using material from a measurement period in the spring of 1994 in Kaepylae, Helsinki. Material from a pre-operational period in the town of Oulu was also available. (author)
A new method for calculation of an air quality index
Energy Technology Data Exchange (ETDEWEB)
Ilvessalo, P [Finnish Meteorological Inst., Helsinki (Finland). Air Quality Dept.
1996-12-31
Air quality measurement programs in Finnish towns have expanded during the last few years. As a result of this it is more and more difficult to make use of all the measured concentration data. Citizens of Finnish towns are nowadays taking more of an interest in the air quality of their surroundings. The need to describe air quality in a simplified form has increased. Air quality indices permit the presentation of air quality data in such a way that prevailing conditions are more easily understandable than when using concentration data as such. Using an air quality index always means that some of the information about concentrations of contaminants in the air will be lost. How much information is possible to extract from a single index number depends on the calculation method. A new method for the calculation of an air quality index has been developed. This index always indicates the overstepping of an air quality guideline level. The calculation of this air quality index is performed using the concentrations of all the contaminants measured. The index gives information both about the prevailing air quality and also the short-term trend. It can also warn about the expected exceeding of guidelines due to one or several contaminants. The new index is especially suitable for the real-time monitoring and notification of air quality values. The behaviour of the index was studied using material from a measurement period in the spring of 1994 in Kaepylae, Helsinki. Material from a pre-operational period in the town of Oulu was also available. (author)
Calculated inclusive neutron production from 400 GeV proton-nucleus collisions
International Nuclear Information System (INIS)
Alsmiller, R.G. Jr.; Alsmiller, F.S.; Hermann, O.W.
1989-08-01
Calculated inclusive neutron production from 400 GeV proton-nucleus collisions is presented and compared with experimental data. Target nuclei H, Be, Cu, and Pb are considered and the comparisons cover the laboratory energy range of 20 to 400 GeV, and angular range 0.7 to 10 mr. Moderately good agreement between the calculated and experimental data is found, but the agreement in the case of Be, Cu and Pb is significantly better than in the case of H. 8 refs., 4 figs
One group neutron flux at a point in a cylindrical reactor cell calculated by Monte Carlo
Energy Technology Data Exchange (ETDEWEB)
Kocic, A [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)
1974-01-15
Mean values of the neutron flux over material regions and the neutron flux at space points in a cylindrical annular cell (one group model) have been calculated by Monte Carlo. The results are compared with those obtained by an improved collision probability method (author)
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Gallego D, E.; Lorente F, A. [Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)
2005-07-01
By means of Monte Carlo methods the spectra of the produced neutrons {sup 252} Cf, {sup 252} Cf/D{sub 2}O, {sup 241} Am Be, {sup 239} Pu Be, {sup 140} La Be, {sup 239} Pu{sup 18}O{sub 2} and {sup 226} Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H{sup *}(10), H{sub p,sIab} (10, 0{sup 0}), E{sub AP} and E{sub ISO}. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of {sup 239} Pu Be and {sup 241} Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)
Calculation of dosimetry parameters for fast neutron radiotherapy
Energy Technology Data Exchange (ETDEWEB)
Wells, A.H.
1978-05-01
A computer simulation of the interactions of 50 MeV d/sup +/ on Be and 42 MeV p/sup +/ on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/.
Calculation of dosimetry parameters for fast neutron radiotherapy
International Nuclear Information System (INIS)
Wells, A.H.
1978-05-01
A computer simulation of the interactions of 50 MeV d + on Be and 42 MeV p + on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/
Analysis and evaluation of critical experiments for validation of neutron transport calculations
International Nuclear Information System (INIS)
Bazzana, S.; Blaumann, H; Marquez Damian, J.I
2009-01-01
The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook. [es
CDFMC: a program that calculates the fixed neutron source distribution for a BWR using Monte Carlo
International Nuclear Information System (INIS)
Gomez T, A.M.; Xolocostli M, J.V.; Palacios H, J.C.
2006-01-01
The three-dimensional neutron flux calculation using the synthesis method, it requires of the determination of the neutron flux in two two-dimensional configurations as well as in an unidimensional one. Most of the standard guides for the neutron flux calculation or fluences in the vessel of a nuclear reactor, make special emphasis in the appropriate calculation of the fixed neutron source that should be provided to the used transport code, with the purpose of finding sufficiently approximated flux values. The reactor core assemblies configuration is based on X Y geometry, however the considered problem is solved in R θ geometry for what is necessary to make an appropriate mapping to find the source term associated to the R θ intervals starting from a source distribution in rectangular coordinates. To develop the CDFMC computer program (Source Distribution calculation using Monte Carlo), it was necessary to develop a theory of independent mapping to those that have been in the literature. The method of meshes overlapping here used, is based on a technique of random points generation, commonly well-known as Monte Carlo technique. Although the 'randomness' of this technique it implies considering errors in the calculations, it is well known that when increasing the number of points randomly generated to measure an area or some other quantity of interest, the precision of the method increases. In the particular case of the CDFMC computer program, the developed technique reaches a good general behavior when it is used a considerably high number of points (bigger or equal to a hundred thousand), with what makes sure errors in the calculations of the order of 1%. (Author)
Air pollution studies in Tianjing city using neutron activation analysis techniques
International Nuclear Information System (INIS)
Ni Bangfa; Tian Weizhi; Nie Nuiling; Wang Pingsheng
1999-01-01
Two sites of airborne sampling from industrial and residential areas were made in Tianjing city during February and June using PM-10 sampler and analyzed by NAA techniques; Comparison of air pollution between urban and rural area in Tianjing city was made using neutron activation analysis techniques and some other data analyzing techniques. (author)
International Nuclear Information System (INIS)
Koch, R.; Findeisen, A.
1986-04-01
The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)
Energy Technology Data Exchange (ETDEWEB)
Trotter, D.E. Gonzalez [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: crowell@tunl.duke.edu; Meneses, F. Salinas [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Tornow, W. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: tornow@tunl.duke.edu; Crowell, A.S.; Howell, C.R. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Schmidt, D. [Physikalisch-Technische Bundesanstalt, D-38116, Braunschweig (Germany); Walter, R.L. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)
2009-02-11
The methods employed and the results obtained from measurements and calculations of the detection efficiency for the neutron detectors used at Triangle Universities Nuclear Laboratory (TUNL) in the simultaneous determination of the {sup 1}S{sub 0} neutron-neutron and neutron-proton scattering lengths a{sub nn} and a{sub np}, respectively, are described. Typical values for the detector efficiency were 0.3. Very good agreement between the different experimental methods and between data and calculation has been obtained in the neutron energy range below E{sub n}=13MeV.
Neutronic calculation of the next fuel elements for the Argonaut reactor
International Nuclear Information System (INIS)
Oliveira, C.R.E.; Brito Aghina, L.O. de
1981-01-01
The best parameters of the next fuel elements of the Argonaut reactor, at IEN (Instituto de Engenharia Nuclear - Brazil), were determined. The next fuel elements will be rods of an uranium mixture (19.98% enriched), graphite and bakelite. The parameters to be determined are: mixture density, percentage of uranium in the mixture, pellet radius, rod material and elements arrangement (step). The calculations routines consisted in the analysis of several steps, using the LEOPARD computer code for cell calculations and RMAT1D for one dimensional spatial calculations (criticality) with four energy groups. Finally a neutronic study of the Argounat reactors present configuration was done, using the HAMMER computer code (cell), the EXTERMINATOR computer code (two-dimensional calculations) and RAMAT1D. (Author) [pt
Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC
International Nuclear Information System (INIS)
West, L.; Akin, B.P.; Lemley, E.C.
1995-01-01
Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations
Dosimetry intercomparisons between fast neutron radiotherapy facilities
International Nuclear Information System (INIS)
Almond, P.R.; Smith, A.R.; Smathers, J.B.; Otte, V.A.
1975-01-01
Neutron dosimetry intercomparisons have been made between M.D. Anderson Hospital and Tumor Institute, Naval Research Laboratory, University of Washington Hospital, and Hammersmith Hospital. The parameters that are measured during these visits are: tissue kerma in air, tissue dose at depth of dose maximum, depth dose, beam profiles, neutron/gamma ratios and photon calibrations of ionization chambers. A preliminary report of these intercomparisons will be given including a comparison of the calculation and statement of tumor doses for each institution
Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility
Energy Technology Data Exchange (ETDEWEB)
Maerker, R.E.; Maudlin, P.J.
1981-02-01
A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.
Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility
International Nuclear Information System (INIS)
Maerker, R.E.; Maudlin, P.J.
1981-02-01
A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment
Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors
2017-03-01
At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.
Energy Technology Data Exchange (ETDEWEB)
Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, Sean [Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Ghani, Zamir [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, Igor [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Milocco, Alberto; Packer, Lee [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, Mario [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)
2017-03-01
At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.
International Nuclear Information System (INIS)
Mayer, R.E.; Patino, N.E.; Florido, P.C.; Gomez, S.E.; Granada, J.R.; Gillette, V.H.
1993-01-01
A 22% efficiency thermal neutron detection system was designed for the investigation of neutron emission from pulsed D 2 O electrolysis. Reasons are discussed for the choice of 10 atm 3 He proportional counters. Optimization calculations carried out through standard reactor code system (AMPX-II) are presented along with construction details and characteristics of the associated electronics. Experimental verification of calculated efficiency and examples of measurements performed with the detector are included. (orig.)
International Nuclear Information System (INIS)
Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.
2014-01-01
Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff
Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy
International Nuclear Information System (INIS)
Krstic, D.; Markovic, V.M.; Jovanovic, Z.; Milenkovic, B.; Nikezic, D.; Atanackovic, J.
2014-01-01
Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. The difference in evaluated dose in cancer and normal lung tissue suggests that BNCT could be applied for the treatment of cancers. The difference in exposure of cancer and healthy tissue can be observed, so the healthy tissue can be spared from damage. An absorbed dose ratio of metastatic tissue-to-the healthy tissue was ∼5. Absorbed dose to all other organs was low when compared with the lung dose. Absorbed dose depth distribution shows that BNC therapy can be very useful in the treatments for tumour. The ratio of the tumour absorbed dose and irradiated healthy tissue absorbed dose was also ∼5. It was seen that an elliptical neutron field was better irradiation choice. (authors)
Calculated microdose spectra for intermediate energy neutrons (1 to 100 keV)
International Nuclear Information System (INIS)
Al-Affan, I.A.M.; Watt, D.E.
1983-01-01
Basic formulae for calculation of energy deposition events due to insiders, starters, stoppers and crossers, using the continuous slowing down approximation have been modified to allow for the enhanced energy deposition in spherical volumes due to elastic scattering interactions which reduce the penetration depth of the charged particle recoils. Energy deposition spectra have been obtained for energies of 1, 10, 50, 100 keV in 0.2 μm and 1 μm tissue-equivalent spheres. From these, frequency and dose distributions in lineal energy and in specific energy density have been calculated. Also calculated for different neutron energies are values of zeta, the energy average of event size, as a function of the diameter of the sensitive site. The structure of the energy event distributions can be interpreted in terms of the basic physics. The effect of the modifications to the basic formulae is to increase the number of energy deposition events due to insiders and to decrease the number of starters, stoppers and crossers. The degree of the effect increases with decreasing neutron energy, increasing sphere size, and the change is most significant for low energy deposition events. (author)
Calculated microdose spectra for intermediate energy neutrons (1 to 100 keV)
Energy Technology Data Exchange (ETDEWEB)
Al-Affan, I.A.M.; Watt, D.E. (Dundee Univ. (UK). Dept. of Medical Biophysics); Colautti, P.; Talpo, G. (Laboratori Nazionali dell' Infn, 35020, Legnaro (Padova) (Italy))
1983-01-01
Basic formulae for calculation of energy deposition events due to insiders, starters, stoppers and crossers, using the continuous slowing down approximation have been modified to allow for the enhanced energy deposition in spherical volumes due to elastic scattering interactions which reduce the penetration depth of the charged particle recoils. Energy deposition spectra have been obtained for energies of 1, 10, 50, 100 keV in 0.2 ..mu..m and 1 ..mu..m tissue-equivalent spheres. From these, frequency and dose distributions in lineal energy and in specific energy density have been calculated. Also calculated for different neutron energies are values of zeta, the energy average of event size, as a function of the diameter of the sensitive site. The structure of the energy event distributions can be interpreted in terms of the basic physics. The effect of the modifications to the basic formulae is to increase the number of energy deposition events due to insiders and to decrease the number of starters, stoppers and crossers. The degree of the effect increases with decreasing neutron energy, increasing sphere size, and the change is most significant for low energy deposition events.
Pellet by pellet neutron flux calculations coupled with nodal expansion method
International Nuclear Information System (INIS)
Aldo, Dall'Osso
2003-01-01
We present a technique whose aim is to replace 2-dimensional pin by pin de-homogenization, currently done in core reactor calculations with the nodal expansion method (NEM), by a 3-dimensional finite difference diffusion calculation. This fine calculation is performed as a zoom in each node taking as boundary conditions the results of the NEM calculations. The size of fine mesh is of the order of a fuel pellet. The coupling between fine and NEM calculations is realised by an albedo like boundary condition. Some examples are presented showing fine neutron flux shape near control rods or assembly grids. Other fine flux behaviour as the thermal flux rise in the fuel near the reflector is emphasised. In general the results show the interest of the method in conditions where the separability of radial and axial directions is not granted. (author)
Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S
2014-08-01
The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.
Comparison of spallation-model calculations with experiment on neutron production in uranium
International Nuclear Information System (INIS)
Kazaritskij, V.D.; Sibirtsev, A.A.; Stepanov, N.V.
1988-01-01
Calculated differential cross-sections for neutrons emitted from a thin uranium target due to 590 MeV protons are compared with experiment. The HETC, HAMLET, D2N2 and EVAP codes were applied. The HAMLET results are most consistent with experiment. 12 refs.; 3 figs
International Nuclear Information System (INIS)
Bernnat, W.; Keinert, J.; Mattes, M.
2004-01-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)
International Nuclear Information System (INIS)
Monteiro, E.
2004-01-01
The aimed of this work consists of evaluating the influence of the secondary contributions of dose (thermal neutrons dose, epithermal neutrons dose, fast neutrons dose and photon dose) in treatment planning with BNCT. MCNP4B Code was used to calculate RBE-Gy doses through the irradiation of the modified Snyder head head phantom.A reduction of the therapeutical gain of monoenergetic neutron beans was observed in non invasive treatments, provoked for the predominance of the fast neutron dose component in the skin, showing that the secondary contributions of dose can contribute more in the direction to raise the dose in the fabric healthy that in the tumor, thus reducing the treatment efficiency. (author)
International Nuclear Information System (INIS)
Pal, Rupali R.; Sathian, Deepa; Jayalakshmi, V.; Chougaonkar, M.P.
2011-01-01
CR-39 is the most sensitive of nuclear track detectors for protons and is recommended as an effective neutron dosimeter because of it's low threshold energy of 100 keV neutrons. The fraction of protons that gives detectable tracks in CR-39 depends on the energy of the proton angle of incidence and etching conditions. As a consequence the registration efficiency of neutrons in the CR-39 plastics used for neutron personnel monitoring is strongly influenced by the direction of radiation incidence. This paper presents the relative response of CR-39 at varying neutron incident angles, for 241 Am-Be neutron source spectra in free air and on ISO phantom, in terms of operational quantities. It is observed that the angular dependence of CR-39 for irradiations in air and on phantom is essentially the same indicating that the phantom does not affect the directional response of CR-39. (author)
Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data
International Nuclear Information System (INIS)
Boado, H.J.; Gho, C.J.; Abbate, M.J.
1981-01-01
Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author) [es
Energy Technology Data Exchange (ETDEWEB)
Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)
2009-07-01
Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)
International Nuclear Information System (INIS)
Mota, F.; Ortiz, C. J.; Vila, R.
2012-01-01
Irradiation Experimental Area of TechnoFusion will emulate the extreme irradiation fusion conditions in materials by means of three ion accelerators: one used for self-implanting heavy ions (Fe, Si, C,...) to emulate the displacement damage induced by fusion neutrons and the other two for light ions (H and He) to emulate the transmutation induced by fusion neutrons. This Laboratory will play an essential role in the selection of functional materials for DEMO reactor since it will allow reproducing the effects of neutron radiation on fusion materials. Ion irradiation produces little or no residual radioactivity, allowing handling of samples without the need for special precautions. Currently, two different methods are used to calculate the primary displacement damage by neutron irradiation or by ion irradiation. On one hand, the displacement damage doses induced by neutrons are calculated considering the NRT model based on the electronic screening theory of Linhard. This methodology is commonly used since 1975. On the other hand, for experimental research community the SRIM code is commonly used to calculate the primary displacement damage dose induced by ion irradiation. Therefore, both methodologies of primary displacement damage calculation have nothing in common. However, if we want to design ion irradiation experiments capable to emulate the neutron fusion effect in materials, it is necessary to develop comparable methodologies of damage calculation for both kinds of radiation. It would allow us to define better the ion irradiation parameters (Ion, current, Ion energy, dose, etc) required to emulate a specific neutron irradiation environment. Therefore, our main objective was to find the way to calculate the primary displacement damage induced by neutron irradiation and by ion irradiation starting from the same point, that is, the PKA spectrum. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on
The calculation of neutron flux using Monte Carlo method
Günay, Mehtap; Bardakçı, Hilal
2017-09-01
In this study, a hybrid reactor system was designed by using 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII.0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.
International Nuclear Information System (INIS)
Perry, R.T.; Wilson, W.B.; Charlton, W.S.
1998-04-01
In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an
Methods of neutron spectrum calculation from measured reaction velocities in SAIPS
International Nuclear Information System (INIS)
Berzonis, M.A.; Bondars, Kh.Ya.
1981-01-01
When a user (physicist) needs to perform calculations, he faces a number of problems: obtaining or generating calculation programs, comparing these programs, generating a library of reference spectra, study of calculated spectra and so on. This means routine work which is duplicated in many laboratories. To help solve these problems a computerized information system called SAIPS has been developed, some aspects of which are dealt with in references. The present paper gives a short description of data input into SAIPS and the basic principles of its utilization. SAIPS is based on the ES 1022 computer controlled by the operational system OS ES version 4.1. It contains the programs needed for unfolding spectra, neutron cross-section and reference spectrum libraries and the software for the main system and for computerized calculations
International Nuclear Information System (INIS)
Jachic, J.
1985-01-01
It is presented the ONEDM neutronic simulator for RZ spatial calculation, two energy groups, aiming at researching and optimization of a low power fast reactor design. The simulator's methodology is based in RZ calculation from radial and axial calculation iteractively coupled and in macroscopic cross sections corrected by power density and asymmetry of the spectrum in the feedback process with phase library for reference neutronic state. The transversal area which are determined by energy groups and material region in the iteration are introduced in the spatial calculation. The simulator efficiency is tested and compared with the CITATION and 2DB codes. The cross sections are generated by 1DX code. (M.C.K.) [pt
Neutron sources and its dosimetric characteristics
International Nuclear Information System (INIS)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A.; Gallego D, E.; Lorente F, A.
2005-01-01
By means of Monte Carlo methods the spectra of the produced neutrons 252 Cf, 252 Cf/D 2 O, 241 Am Be, 239 Pu Be, 140 La Be, 239 Pu 18 O 2 and 226 Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H * (10), H p,sIab (10, 0 0 ), E AP and E ISO . During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of 239 Pu Be and 241 Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)
Energy Technology Data Exchange (ETDEWEB)
Villescas, G.; Corchon, F.
2013-07-01
he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.
Kasesaz, Y; Khalafi, H; Rahmani, F
2013-12-01
Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45 MeV neutrons that are produced in the D-D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70 cm Fluental as a moderator, 30 cm Pb as a reflector, 2mm (6)Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method. Copyright © 2013 Elsevier Ltd. All rights reserved.
Neutron-deuteron scattering calculations with W-matrix representation of the two-body input
International Nuclear Information System (INIS)
Bartnik, E.A.; Haberzettl, H.; Januschke, T.; Kerwath, U.; Sandhas, W.
1987-05-01
Employing the W-matrix representation of the partial-wave T matrix introduced by Bartnik, Haberzettl, and Sandhas, we show for the example of the Malfliet-Tjon potentials I and III that the single-term separable part of the W-matrix representation, when used as input in three-nucleon neutron-deuteron scattering calculations, is fully capable of reproducing the exact results obtained by Kloet and Tjon. This approximate two-body input not only satisfies the two-body off-shell unitarity relation but, moreover, it also contains a parameter which may be used in optimizing the three-body data. We present numerical evidence that there exists a variational (minimum) principle for the determination of the three-body binding energy which allows one to choose this parameter also in the absence of an exact reference calculation. Our results for neutron-deuteron scattering show that it is precisely this choice of the parameter which provides optimal scattering data. We conclude that the W-matrix approach, despite its simplicity, is a remarkably efficient tool for high-quality three-nucleon calculations. (orig.)
Data and software for calculating neutron spectra from measured reaction rates
International Nuclear Information System (INIS)
Berzonis, M.A.; Bonbars, Kh.Ya.
1981-01-01
The information system SAIPS is presented, which allows the automated calculation of neutron spectra and the use of cross section libraries on EC type computers. The following programmes can be applied: SAND II, WINDOWS, CRYSTAL BALL, RFSP JUEL, etc. The system includes both cross section libraries established by means of the code mentioned and libraries recommended by several laboratories. (author)
Energy Technology Data Exchange (ETDEWEB)
Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)
2004-03-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)
The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel
International Nuclear Information System (INIS)
Gurevich, M.I.; Brizgalov, V.I.
1994-01-01
This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig
Energy Technology Data Exchange (ETDEWEB)
Gerasimenko, B.F. [V.G. Khlopin Radium Inst., Saint Peterburg (Russian Federation)
1997-03-01
The calculations of integral spectra of prompt neutrons of spontaneous fission of {sup 244}Cm and {sup 246}Cm were carried out. The calculations were done by the Statistical Computer Code Complex SCOFIN applying the Hauser-Feschbach method as applied to the description of the de-excitation of excited fission fragments by means of neutron emission. The emission of dipole gamma-quanta from these fragments was considered as a competing process. The average excitation energy of a fragment was calculated by two-spheroidal model of tangent fragments. The density of levels in an excited fragment was calculated by the Fermi-gas model. The quite satisfactory agreement was reached between theoretical and experimental results obtained in frames of Project measurements. The calculated values of average multiplicities of neutron number were 2,746 for {sup 244}Cm and 2,927 for {sup 246}Cm that was in a good accordance with published experimental figures. (author)
Neutron and photon transport calculations in fusion system. 2
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1998-03-01
On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)
Calculated neutron spectrum from 800-MeV protons incident on a copper beam stop
International Nuclear Information System (INIS)
Perry, D.G.
1975-10-01
A Monte Carlo calculation was performed to obtain the neutron spectrum generated by 800-MeV protons incident on the LAMPF main copper beam stop. The total flux is calculated to be of the order of 10 13 n/cm 2 -sec-mA at full-beam intensity of 1 mA, with flux spectra calculated for angles of 20 0 , 30 0 , 60 0 , 90 0 , 120 0 , and 150 0 . (auth)
Jansky, Bohumil; Rejchrt, Jiri; Novak, Evzen; Losa, Evzen; Blokhin, Anatoly I.; Mitenkova, Elena
2017-09-01
The leakage neutron spectra measurements have been done on benchmark spherical assemblies - iron spheres with diameter of 20, 30, 50 and 100 cm. The Cf-252 neutron source was placed into the centre of iron sphere. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional counters with diameter of 4 cm and with pressure of 400 and 1000 kPa. The neutron energy range of spectrometer is from 0.1 to 1.3 MeV. This energy interval represents about 85 % of all leakage neutrons from Fe sphere of diameter 50 cm and about of 74% for Fe sphere of diameter 100 cm. The adequate MCNP neutron spectra calculations based on data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 were done. Two calculations were done with CIELO library. The first one used data for all Fe-isotopes from CIELO and the second one (CIELO-56) used only Fe-56 data from CIELO and data for other Fe isotopes were from ENDF/B-VII.1. The energy structure used for calculations and measurements was 40 gpd (groups per decade) and 200 gpd. Structure 200 gpd represents lethargy step about of 1%. This relatively fine energy structure enables to analyze the Fe resonance neutron energy structure. The evaluated cross section data of Fe were validated on comparisons between the calculated and experimental spectra.
International Nuclear Information System (INIS)
Chernyshov, V.S.; Kalebin, S.M.; Ivanov, R.N.; Pershukov, A.P.; Artamonov, V.S.
1989-01-01
Optimal sizes of the proton detector focusing system used in the experiment on determination of life-time of a free neutron are calculated. Angles of proton incidence on the plane of the detector window don't exceed 27 deg. Distances which protons fly along the neutron beam axis before they appear in the detector is less 4.3 cm. Analysis of proton trajectories an investigation of focusing system stability during transverse oscillations of a neutron beam permits to choose necessary width of the proton counter window by means of which all protons run to recording volume. 10 refs.; 8 figs
MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data
International Nuclear Information System (INIS)
2001-01-01
1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation
International Nuclear Information System (INIS)
Khoshnoodi, M.; Sohrabi, M.
1997-01-01
A new expression modifying the inverse square law for determination of neutron fluences from spherical moderator neutron sources is reported. The formalism is based on the neutron fluence at a point outside the moderator as the summation of fluxes of two groups of neutrons: direct neutrons from the central region of the moderator, and moderated neutrons which, to a first approximation, are scattered from the outermost layers of the spherical moderator. The expression has been further developed for spherical neutron measuring devices with an appropriate geometry factor which corrects the reading of the device for non-uniform irradiation of the detector. The combination of the new fluence function and those of the air and room scattered components introduce a calibration model. The fluence relationship obtained for moderated sources may conveniently be used for calculating the more rapid change of neutron dose at close distances than that which is based on the inverse square dependence. (author)
Adjusted neutron spectra of STEK cores for reactivity calculations
International Nuclear Information System (INIS)
Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.
1978-02-01
Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK
Nie, Y. B.; Ruan, X. C.; Ren, J.; Zhang, S.; Han, R.; Bao, J.; Huang, H. X.; Ding, Y. Y.; Wu, H. C.; Liu, P.; Zhou, Z. Y.
2017-09-01
In order to make benchmark validation of the nuclear data for gallium (Ga), tungsten (W) and beryllium (Be) in existing modern evaluated nuclear data files, neutron leakage spectra in the range from 0.8 to 15 MeV from slab samples were measured by time-of-flight technique with a BC501 scintillation detector. The measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source. The thicknesses of the slabs were 0.5 to 2.5 mean free path for 14.8 MeV neutrons, and the measured angles were chosen to be 60∘ and 120∘. The measured spectra were compared with those calculated by the continuous energy Monte-Carlo transport code MCNP, using the data from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 nuclear data files, the comparison between the experimental and calculated results show that: The results from all three libraries significantly underestimate the cross section in energy range of 10-13 MeV for Ga; For W, the calculated spectra using data from CENDL-3.1 and JENDL-4.0 libraries show larger discrepancies with the measured ones, especially around 8.5-13.5 MeV; and for Be, all the libraries led to underestimation below 3 MeV at 120∘.
Fuel management and core design code systems for pressurized water reactor neutronic calculations
International Nuclear Information System (INIS)
Ahnert, C.; Arayones, J.M.
1985-01-01
A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions
Energy Technology Data Exchange (ETDEWEB)
Slater, C.O.
1990-07-01
Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.
Design and spectrum calculation of 4H-SiC thermal neutron detectors using FLUKA and TCAD
Huang, Haili; Tang, Xiaoyan; Guo, Hui; Zhang, Yimen; Zhang, Yimeng; Zhang, Yuming
2016-10-01
SiC is a promising material for neutron detection in a harsh environment due to its wide band gap, high displacement threshold energy and high thermal conductivity. To increase the detection efficiency of SiC, a converter such as 6LiF or 10B is introduced. In this paper, pulse-height spectra of a PIN diode with a 6LiF conversion layer exposed to thermal neutrons (0.026 eV) are calculated using TCAD and Monte Carlo simulations. First, the conversion efficiency of a thermal neutron with respect to the thickness of 6LiF was calculated by using a FLUKA code, and a maximal efficiency of approximately 5% was achieved. Next, the energy distributions of both 3H and α induced by the 6LiF reaction according to different ranges of emission angle are analyzed. Subsequently, transient pulses generated by the bombardment of single 3H or α-particles are calculated. Finally, pulse height spectra are obtained with a detector efficiency of 4.53%. Comparisons of the simulated result with the experimental data are also presented, and the calculated spectrum shows an acceptable similarity to the experimental data. This work would be useful for radiation-sensing applications, especially for SiC detector design.
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
Calculation analysis of Wims/D4-Batan-2DIFF neutron spectrum on RSG-GAS with cadmium ratio
International Nuclear Information System (INIS)
Radianti, I.B.; Zuhair; Hamzah, A.
1998-01-01
The calculation analysis of WIMS/D4-BATAN-2DIFF neutron spectrum was performed by comparison the calculation result of cadmium ratio with the experiment result on CIP, IP2, IP3 and IP4 irradiation positions of RSG GAS tenth core. The foils of Au, Mn and Co were used for determination of the measured and calculated cadmium ratios. Spectrum calculation was done in 69 energy group with 541 energy group (till 10 MeV) cross section of foil absorption reaction. The difference values between cadmium ratio calculation and experiment result for all cases were in interval of 11.0%-26.3% which are out of measurement deviation range. From these result, it concluded that the use of WIM /D4 in generating group constant is not sufficient to obtain the neutron spectrum, especially for non-fuel region
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
New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code
Energy Technology Data Exchange (ETDEWEB)
Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)
2016-10-15
The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.
Calculation principles of humid air in a reversed Brayton cycle
Energy Technology Data Exchange (ETDEWEB)
Backman, J [Lappeenranta Univ. of Technology (Finland). Dept. of Energy Technology
1998-12-31
The article presents a calculation method for reversed Brayton cycle that uses humid air as working medium. The reversed Brayton cycle can be employed as an air dryer, a heat pump or a refrigerating machine. In this research the use of humid air as a working fluid has an environmental advantage, as well. In this method especially the expansion process in the turbine is important because of the condensation of the water vapour in the humid air. This physical phenomena can have significant effects on the level of performance of the application. The expansion process differs physically from the compression process, when the water vapour in the humid air begins to condensate. In the thermodynamic equilibrium of the flow, the water vapour pressure in humid air cannot exceed the pressure of saturated water vapour in corresponding temperature. Expansion calculation during operation around the saturation zone is based on a quasistatic expansion, in which the system after the turbine is in thermodynamical equilibrium. The state parameters are at every moment defined by the equation of state, and there is no supercooling in the vapour. Following simplifications are used in the calculations: The system is assumed to be adiabatic. This means that there is no heat transfer to the surroundings. This is a common practice, when the temperature differences are moderate as here; The power of the cooling is omitted. The cooling construction is very dependent on the machine and the distribution of the losses; The flow is assumed to be one-dimensional, steady-state and homogenous. The water vapour condensing in the turbine can cause errors, but the errors are mainly included in the efficiency calculation. (author) 11 refs.
Calculation principles of humid air in a reversed Brayton cycle
Energy Technology Data Exchange (ETDEWEB)
Backman, J. [Lappeenranta Univ. of Technology (Finland). Dept. of Energy Technology
1997-12-31
The article presents a calculation method for reversed Brayton cycle that uses humid air as working medium. The reversed Brayton cycle can be employed as an air dryer, a heat pump or a refrigerating machine. In this research the use of humid air as a working fluid has an environmental advantage, as well. In this method especially the expansion process in the turbine is important because of the condensation of the water vapour in the humid air. This physical phenomena can have significant effects on the level of performance of the application. The expansion process differs physically from the compression process, when the water vapour in the humid air begins to condensate. In the thermodynamic equilibrium of the flow, the water vapour pressure in humid air cannot exceed the pressure of saturated water vapour in corresponding temperature. Expansion calculation during operation around the saturation zone is based on a quasistatic expansion, in which the system after the turbine is in thermodynamical equilibrium. The state parameters are at every moment defined by the equation of state, and there is no supercooling in the vapour. Following simplifications are used in the calculations: The system is assumed to be adiabatic. This means that there is no heat transfer to the surroundings. This is a common practice, when the temperature differences are moderate as here; The power of the cooling is omitted. The cooling construction is very dependent on the machine and the distribution of the losses; The flow is assumed to be one-dimensional, steady-state and homogenous. The water vapour condensing in the turbine can cause errors, but the errors are mainly included in the efficiency calculation. (author) 11 refs.
Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)
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.
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.
International Nuclear Information System (INIS)
Barz, H.U.; Bertram, W.
1992-01-01
Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Krueger, Thomas
2016-10-19
The physics of neutron-rich systems is of great interest in nuclear and astrophysics. Precise knowledge of the properties of neutron-rich nuclei is crucial for understanding the synthesis of heavy elements. Infinite neutron matter determines properties of neutron stars, a final stage of heavy stars after a core-collapse supernova. It also provides a unique theoretical laboratory for nuclear forces. Strong interactions are determined by quantum chromodynamics (QCD). However, QCD is non-perturbative at low energies and one presently cannot directly calculate nuclear forces from it. Chiral effective field theory circumvents these problems and connects the symmetries of QCD to nuclear interactions. It naturally and systematically includes many-nucleon forces and gives access to uncertainty estimates. We use chiral interactions throughout all calculation in this thesis. Neutron stars are very extreme objects. The densities in their interior greatly exceed those in nuclei. The exact composition and properties of neutron stars is still unclear but they consist mainly of neutrons. One can explore neutron stars theoretically with calculations of neutron matter. In the inner core of neutron stars exist very high densities and thus maybe exotic phases of matter. To investigate whether there exists a phase transition to such phases even at moderate densities we study the chiral condensate in neutron matter, the order parameter of chiral symmetry breaking, and find no evidence for a phase transition at nuclear densities. We also calculate the more extreme system of spin-polarised neutron matter. With this we address the question whether there exists such a polarised phase in neutron stars and also provide a benchmark system for lattice QCD. We find spin-polarised neutron matter to be an almost non-interacting Fermi gas. To understand the cooling of neutron stars neutron pairing is of great importance. Due to the high densities especially triplet pairing is of interest. We
The adaptation of methods in multilayer optics for the calculation of specular neutron reflection
International Nuclear Information System (INIS)
Penfold, J.
1988-10-01
The adaptation of standard methods in multilayer optics to the calculation of specular neutron reflection is described. Their application is illustrated with examples which include a glass optical flat and a deuterated Langmuir-Blodgett film. (author)
International Nuclear Information System (INIS)
Berzonis, M.A.; Bondars, H.Ya.
1981-08-01
A brief description of the SAIPS software and the basic principles of its application is given. SAIPS contains programs needed to unfold spectra, libraries of neutron cross sections and reference spectra, and software for automatic calculation and for system maintenance. SAIPS offers the possibility of determining the reliability of an unfolded neutron spectrum and of planning measurements and calculations by varying different factors: the errors in the reaction rates, the errors in the cross sections used, the detector assembly, the unfolding programs, etc. SAIPS runs on the ES 1022 computer
Calculations of the spectra of fast neutrons in iron spheres using the vitamin-C file
International Nuclear Information System (INIS)
Ahmed, F.; Aizawa, O.; Kadotani, H.
1984-01-01
Steady-state space-dependent fast neutron angular and scalar spectra and total flux in various iron spheres have been calculated using the one-dimensional discrete ordinate transport code ANISN and Vitamin-C nuclear data file. The results have been used to study the question of establishment of equilibrium and of an associated fast neutron diffusion length in iron. The authors find that true equilibrium conditions are not established even inside a 3-m-radius iron sphere. However, from the study of spatial decay of total flux, one can obtain the value of the fast neutron diffusion length in iron, which in the present case is found to be 24.4 cm
Methods and codes for neutronic calculations of the MARIA research reactor
International Nuclear Information System (INIS)
Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.
1998-01-01
The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)
International Nuclear Information System (INIS)
Keney, G.S.
1981-08-01
A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV
International Nuclear Information System (INIS)
Gardner, D.G.
1975-01-01
A large amount of cross section and spectral information for neutron-induced reactions will be required for the CTR design program. To undertake to provide the required data through a purely experimental measurement program alone may not be the most efficient way of attacking the problem. It is suggested that a preliminary theoretical calculation be made of all relevant reactions on the dozen or so elements that now seem to comprise the inventory of possible construction materials to find out which are actually important, and over what energy ranges they are important. A number of computer codes for calculating cross sections for neutron induced reactions have been evaluated and extended. These will be described and examples will be given of various types of calculations of interest to the CTR program. (U.S.)
OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization
Energy Technology Data Exchange (ETDEWEB)
Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)
2017-06-15
Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.
Goldhagen, P E; Kniss, T; Reginatto, M; Singleterry, R C; Van Steveninck, W; Wilson, J W
2002-01-01
Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from galactic cosmic radiation. Crews of future high-speed commercial aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the atmospheric ionizing radiation (AIR) project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (t...
International Nuclear Information System (INIS)
Benmansour, L.
1992-01-01
The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)
Neutron spectra produced by moderating an isotopic neutron source
International Nuclear Information System (INIS)
Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene
2001-01-01
A Monte Carlo study has been carried out to determine the neutron spectra produced by an isotopic neutron source inserted in moderating media. Most devices used for radiation protection have a response strongly dependent on neutron energy. ISO recommends several neutron sources and monoenergetic neutron radiations, but actual working situations have broad spectral neutron distributions extending from thermal to MeV energies, for instance, near nuclear power plants, medical applications accelerators and cosmic neutrons. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices in neutron spectra which are nearly like those met in practice. In order to complete the range of neutron calibrating sources, it seems useful to develop several wide spectral distributions representative of typical spectra down to thermal energies. The aim of this investigation was to use an isotopic neutron source in different moderating media to reproduce some of the neutron fields found in practice. MCNP code has been used during calculations, in these a 239PuBe neutron source was inserted in H2O, D2O and polyethylene moderators. Moderators were modeled as spheres and cylinders of different sizes. In the case of cylindrical geometry the anisotropy of resulting neutron spectra was calculated from 0 to 2 . From neutron spectra dosimetric features were calculated. MCNP calculations were validated by measuring the neutron spectra of a 239PuBe neutron source inserted in a H2O cylindrical moderator. The measurements were carried out with a multisphere neutron spectrometer with a 6LiI(Eu) scintillator. From the measurements the neutron spectrum was unfolded using the BUNKIUT code and the UTA4 response matrix. Some of the moderators with the source produce a neutron spectrum close to spectra found in actual applications, then can be used during the calibration of radiation protection devices
International Nuclear Information System (INIS)
Ku, L.P.; Price, W.G. Jr.
1977-08-01
The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model
Analysis of fusion neutronics calculations and appraisal of UW cross-section library
International Nuclear Information System (INIS)
Xie Jianping; Li Xingzhong; Ying Chuntong
1989-01-01
A series of calculations for different cases (especially for the values of tritium breeding ratio T, and the fuel breeding ratio F in the blanket of a hybrid reactor) were carried out by using ANISN program and UW cross-section library. The comparison with other results in China and abroad kalso was done. It was shownwn that the installation and execution of ANISN program on ELXSI machine at Tsinghua University are successful, and the UW cross-section library is reliable. It may be used for fusion neutronics calculation in the future. The paper also points out that the difference between the calculations and by the authors are due to jthe different in cross-section data used
Energy Technology Data Exchange (ETDEWEB)
Holmqvist, B; Wiedling, T
1969-06-15
Neutron elastic scattering cross section measurements have been going on for a long period at the Studsvik Van de Graaff laboratory. The cross sections of a range of elements have been investigated in the energy interval 1.5 to 8 MeV. The experimental data have been compared with cross sections calculated with the optical model when using a local nuclear potential.
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)
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)
Calculations of air cooler for new subsonic wind tunnel
Rtishcheva, A. S.
2017-10-01
As part of the component development of TsAGI’s new subsonic wind tunnel where the air flow velocity in the closed test section is up to 160 m/sec hydraulic and thermal characteristics of air cooler are calculated. The air cooler is one of the most important components due to its highest hydraulic resistance in the whole wind tunnel design. It is important to minimize its hydraulic resistance to ensure the energy efficiency of wind tunnel fans and the cost-cutting of tests. On the other hand the air cooler is to assure the efficient cooling of air flow in such a manner as to maintain the temperature below 40 °C for seamless operation of measuring equipment. Therefore the relevance of this project is driven by the need to develop the air cooler that would demonstrate low hydraulic resistance of air and high thermal effectiveness of heat exchanging surfaces; insofar as the cooling section must be given up per unit time with the amount of heat Q=30 MW according to preliminary evaluations. On basis of calculation research some variants of air cooler designs are proposed including elliptical tubes, round tubes, and lateral plate-like fins. These designs differ by the number of tubes and plates, geometrical characteristics and the material of finned surfaces (aluminium or cooper). Due to the choice of component configurations a high thermal effectiveness is achieved for finned surfaces. The obtained results form the basis of R&D support in designing the new subsonic wind tunnel.
Q resolution calculation of small angle neutron scattering spectrometer and analysis of form factor
International Nuclear Information System (INIS)
Chen Liang; Peng Mei; Wang Yan; Sun Liangwei; Chen Bo
2011-01-01
The calculational methods of Small Angle Neutron Scattering (SANS) spectrometer Q resolution function and its correlative Q standard difference were introduced. The effects of Q standard difference were analysed with the geometry lay out of spectrometer and the spread of neutron wavelength. The one dimension Q resolution Gaussian function were analysed. The form factor curve of ideal solid sphere and two different instrument arrangement parameter was convoluted respectively and the different smearing curve of form factor was obtained. The combination of using the Q resolution function to more accurately analysis SANS data. (authors)
Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation
International Nuclear Information System (INIS)
Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.
2009-01-01
The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)
Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation
International Nuclear Information System (INIS)
Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.
2009-01-01
The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)
An optimized ultra-fine energy group structure for neutron transport calculations
International Nuclear Information System (INIS)
Huria, Harish; Ouisloumen, Mohamed
2008-01-01
This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)
Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel
2014-01-01
High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.
Neutronic calculations for the conceptual design of an in-reactor solid breeder experiment, TRIO-01
International Nuclear Information System (INIS)
Childs, R.L.; Gabriel, T.A.; Lillie, R.A.
1981-03-01
Neutronics calculations have been performed to obtain tritium production and heat generation rates for the irradiation of solid tritium breeding materials in the Oak Ridge Research Reactor (ORR). Two breeder materials, Li 2 O and LiAlO 2 , were considered. Burnup calculations were performed to estimate the amount of 6 Li present as a function of time
International Nuclear Information System (INIS)
Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.
1976-03-01
The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)
International Nuclear Information System (INIS)
Krotov, A.D.; Son'ko, A.V.
2009-01-01
Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru
Thermal neutron dose calculations in a brain phantom from 7Li(p,n) reaction based BNCT setup
International Nuclear Information System (INIS)
Elshahat, B.A.; Naqvi, A.A.; Maalej, N.; Abdallah, Khalid
2006-01-01
Monte Carlo simulations were carried out to calculate neutron dose in a brain phantom from a 7 Li(p,n) reaction based setup utilizing a high density polyethylene moderator with graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal /(fast +thermal) neutron intensity ratio as a function of geometric parameters of the setup. Results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated Peak Therapeutic Ratio for the setup was found to be 2.15. With further improvement in the moderator design and brain phantom irradiation arrangement, the setup capabilities can be improved to reach further deep-seated tumor. (author)
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
Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutrons
International Nuclear Information System (INIS)
Stevenson, B.
1985-01-01
Literature in the area of anisotropic neutron scattering is by no means lacking. Attention, however, is usually devoted to solution of some particular neutron transport problem and the model employed is at best approximate. The present approach to the problem in general is classically exact and may be of some particular value to individuals seeking exact numerical results in transport calculations. For attempts neutrons originally directed toward the unit vector Omega, it attempts the evaluation of p(theta'), defined such that p(theta') d theta' is that fraction of scattered neutrons that emerges in the vicinity of a cone i.e., having been scattered to between angles theta' and theta' + d theta' with the axis of preferred orientation i; Omega makes an angle theta with i. The relative simplicity of the final form of the solution for hydrogen, in spite of the complicated nature of the limits involved, is a trade-off that truly is not necessary. The exact general solution presented here in integral form, has exceedingly simple limits, i.e., 0 ≤ theta' ≤ π regardless of the material involved; but the form of the final solution is extraordinarily complicated
International Nuclear Information System (INIS)
Aufiero, Manuele; Brovchenko, Mariya; Cammi, Antonio; Clifford, Ivor; Geoffroy, Olivier; Heuer, Daniel; Laureau, Axel; Losa, Mario; Luzzi, Lelio; Merle-Lucotte, Elsa; Ricotti, Marco E.; Rouch, Hervé
2014-01-01
Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for β eff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (β eff ) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions β eff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed
International Nuclear Information System (INIS)
Artioli, C.; Cignani, B.; Dominici, G.; Frattini, F.; Tavoni, R.
1980-01-01
A number of critical configurations installed in the Masurca experimental reactor of the French Commissariat a l'energie atomique (CEA) in Cadarache were used to simulate, by means of certain approximations, the neutron properties of the core of the PEC reactor. As the amount of plutonium available was insufficient, the Pecore experiment was carried out by using two different fuel zones: one with mixed U and Pu oxide, the other with enriched U oxide. However, an effort was made to keep the two zones similar as far as the neutron energy spectrum was concerned. The first interpretation of Pecore was carried out by the CEA with the assistance of staff of the Italian National Nuclear Energy Committee (CNEN), on the basis of French nuclear data and calculation methods. The paper describes the interpretation made in Italy on the basis of the codes and cross-sections used for the PEC neutron calculations, so that use could be made in this project of the discrepancies between calculation and experiment evaluated for Pecore. The characteristics compared were: critical mass, negative reactivity introduced by the control rods, ratios of reaction rates, efficiency of the nickel-base reflector and the axial and radial power curves. By way of a general conclusion, it can be stated that the interpretations of the CEA and the CNEN are in good agreement; the differences which they show are nearly always within the margin of experimental error. In addition to the main results of the Pecore experiment, the paper deals with certain aspects of: the PEC neutron calculation, from the definition of fuel enrichment to the calculation of the power level reached in each element, and from the updating of reactivity coefficients to the formulation of refuelling strategy. For some sectors, the results of the Pecore experiment had to be taken into account. (author)
Linear cascade calculations of matrix due to neutron-induced nuclear reactions
International Nuclear Information System (INIS)
Avila, Ricardo E
2000-01-01
A method is developed to calculate the total number of displacements created by energetic particles resulting from neutron-induced nuclear reactions. The method is specifically conceived to calculate the damage in lithium ceramics by the 6L i(n, α)T reaction. The damage created by any particle is related to that caused by atoms from the matrix recoiling after collision with the primary particle. An integral equation for that self-damage is solved by interactions, using the magic stopping powers of Ziegler, Biersack and Littmark. A projectile-substrate dependent Kinchin-Pease model is proposed, giving and analytic approximation to the total damage as a function of the initial particle energy (au)
Neutron flux calculation by means of Monte Carlo methods
International Nuclear Information System (INIS)
Barz, H.U.; Eichhorn, M.
1988-01-01
In this report a survey of modern neutron flux calculation procedures by means of Monte Carlo methods is given. Due to the progress in the development of variance reduction techniques and the improvements of computational techniques this method is of increasing importance. The basic ideas in application of Monte Carlo methods are briefly outlined. In more detail various possibilities of non-analog games and estimation procedures are presented, problems in the field of optimizing the variance reduction techniques are discussed. In the last part some important international Monte Carlo codes and own codes of the authors are listed and special applications are described. (author)
Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR
International Nuclear Information System (INIS)
Martinez C, E.
2011-01-01
One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm 2 s, at a height H 4 (239.07 cm) and angle 32.236 o in the core shroud and 4.00 E + 09 n/cm 2 s at a height H 4 and angle 35.27 o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)
Distributions of neutron and gamma doses in phantom under a mixed field
International Nuclear Information System (INIS)
Beraud-Sudreau, E.
1982-06-01
A calculation program, based on Monte Carlo method, allowed to estimate the absorbed doses relatives to the reactor primary radiation, in a water cubic phantom and in cylindrical phantoms modelized from tissue compositions. This calculation is a theoretical approach of gamma and neutron dose gradient study in an animal phantom. PIN junction dosimetric characteristics have been studied experimentally. Air and water phantom radiation doses measured by PIN junction and lithium 7 fluoride, in reactor field have been compared to doses given by dosimetry classical techniques as tissue equivalent plastic and aluminium ionization chambers. Dosimeter responses have been employed to evaluate neutron and gamma doses in plastinaut (tissue equivalent plastic) and animal (piglet). Dose repartition in the piglet bone medulla has been also determined. This work has been completed by comparisons with Doerschell, Dousset and Brown results and by neutron dose calculations; the dose distribution related to lineic energy transfer in Auxier phantom has been also calculated [fr
A study on calculation method for mechanical impedance of air spring
International Nuclear Information System (INIS)
Changgeng, Shuai; Penghui, Li; Rustighi, Emiliano
2016-01-01
This paper proposes an approximate analytic method of obtaining the mechanical impedance of air spring. The sound pressure distribution in cylindrical air spring is calculated based on the linear air wave theory. The influences of different boundary conditions on the acoustic pressure field distribution in cylindrical air spring are analysed. A 1-order ordinary differential matrix equation for the state vector of revolutionary shells under internal pressure is derived based on the non-moment theory of elastic thin shell. Referring to the transfer matrix method, a kind of expanded homogeneous capacity high precision integration method is introduced to solve the non-homogeneous matrix differential equation. Combined the solved stress field of shell with the calculated sound pressure field in air spring under the displacement harmonic excitation, the approximate analytical expression of the input and transfer mechanical impedance for the air spring can be achieved. The numerical simulation with the Comsol Multiphysics software verifies the correctness of theoretical analysis result. (paper)
Maucec, M
2005-01-01
Monte Carlo simulations for nuclear logging applications are considered to be highly demanding transport problems. In this paper, the implementation of weight-window variance reduction schemes in a 'manual' fashion to improve the efficiency of calculations for a neutron logging tool is presented.
International Nuclear Information System (INIS)
Wasastjerna, F.; Lux, I.
1980-03-01
A transmission probability method implemented in the program TPHEX is described. This program was developed for the calculation of neutron flux distributions in hexagonal light water reactor fuel assemblies. The accuracy appears to be superior to diffusion theory, and the computation time is shorter than that of the collision probability method. (author)
Energy Technology Data Exchange (ETDEWEB)
Al-abyad, Mogahed; Mohamed, Gehan Y. [Atomic Energy Authority, Cairo (Egypt). Experimental Nuclear Physics Dept.
2017-08-01
Neutron capture cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the reaction {sup 186}W(n,γ){sup 187}W were measured experimentally using the research reactor (ETRR-2) and an Am-Be neutron source, also calculated using TALYS-1.6 code. The present results of σ{sub 0} are (39.08±2.6, 38.75±0.98 and 38.33 barn) and I{sub 0} are (418.5±74, 439.3±36 and 445.5 barn) by using the reactor, neutron source and TALYS-1.6, respectively. The present results are in acceptable agreement with most of the previous experimental and evaluated data as well as the theoretical calculations.
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
A program for calculating group constants on the basis of libraries of evaluated neutron data
International Nuclear Information System (INIS)
Sinitsa, V.V.
1987-01-01
The GRUKON program is designed for processing libraries of evaluated neutron data into group and fine-group (having some 300 groups) microscopic constants. In structure it is a package of applications programs with three basic components: a monitor, a command language and a library of functional modules. The first operative version of the package was restricted to obtaining mid-group non-block cross-sections from evaluated neutron data libraries in the ENDF/B format. This was then used to process other libraries. In the next two versions, cross-section table conversion modules and self-shielding factor calculation modules, respectively, were added to the functions already in the package. Currently, a fourth version of the GRUKON applications program package, for calculation of sub-group parameters, is under preparation. (author)
A comparison of Nodal methods in neutron diffusion calculations
Energy Technology Data Exchange (ETDEWEB)
Tavron, Barak [Israel Electric Company, Haifa (Israel) Nuclear Engineering Dept. Research and Development Div.
1996-12-01
The nuclear engineering department at IEC uses in the reactor analysis three neutron diffusion codes based on nodal methods. The codes, GNOMERl, ADMARC2 and NOXER3 solve the neutron diffusion equation to obtain flux and power distributions in the core. The resulting flux distributions are used for the furl cycle analysis and for fuel reload optimization. This work presents a comparison of the various nodal methods employed in the above codes. Nodal methods (also called Coarse-mesh methods) have been designed to solve problems that contain relatively coarse areas of homogeneous composition. In the nodal method parts of the equation that present the state in the homogeneous area are solved analytically while, according to various assumptions and continuity requirements, a general solution is sought out. Thus efficiency of the method for this kind of problems, is very high compared with the finite element and finite difference methods. On the other hand, using this method one can get only approximate information about the node vicinity (or coarse-mesh area, usually a feel assembly of a 20 cm size). These characteristics of the nodal method make it suitable for feel cycle analysis and reload optimization. This analysis requires many subsequent calculations of the flux and power distributions for the feel assemblies while there is no need for detailed distribution within the assembly. For obtaining detailed distribution within the assembly methods of power reconstruction may be applied. However homogenization of feel assembly properties, required for the nodal method, may cause difficulties when applied to fuel assemblies with many absorber rods, due to exciting strong neutron properties heterogeneity within the assembly. (author).
Dosimetry of atmospheric neutrons: aircrew dosimetry and therapeutic applications
International Nuclear Information System (INIS)
Tatje, Jennifer
2008-01-01
This trainee-ship reports addresses the quantification of the dose received, in real time, by air-crews during commercial flights. Thus, the author first presents the radiative environment which surrounds people and components, and the possible consequences on this exposure. The different parameters influencing the received dose are developed and discussed. The author then describes the French SIEVERT calculation code which is used by all air companies. He also gives a detailed attention to the legal framework regarding radiation protection. In the next part, the author discusses the use of neutrons applied for therapeutic purposes, and their biological effects such as the bystander effect and the radio-sensitivity to low doses. He describes what is a cancer, and presents a therapeutic technique, the Boron Neutron Capture Therapy (BNCT), which is indicated for a certain type of brain cancer, the glioblastoma. The third part proposes an overview of the state-of-the-art of neutron dosimeters, and more particularly those doped with boron, for dose measurement
Characterisation of air particulate matter in Klang Valley by neutron activation analysis technique
International Nuclear Information System (INIS)
Mohd Suhaimi Hamzah; Shamsiah Abd Rahman; Mohd Khalid Matori; Abd Khalik Wood
2000-01-01
Air particulate matter is known to affect human health, impairs visibility and can cause climate change. Study on air particulate matter in term of particle size and chemical contents is very important to indicate the quality of air in a sampling area. Information on concentration of important constituents in air particles can be used to identify some of emission sources which contribute to the pollution problem. The data collected may also be, used as a basis to design a strategy in order to overcome the air pollution problem in the area. The study involved sampling of air dust at two stations, one in Bangi and the other in Kuala Lumpur using Gent Stack Sampler units. Each sampler capable of collecting air particle sizes smaller than 2.5 micron (PM 2.5) and between 2.5 - O micron on two different filters simultaneously. The filters were measured for their mass, elemental carbon and elemental concentrations using analytical equipment or techniques including reflectometer and Neutron Activation Analysis. The results of analysis on samples collected in 1997-1998 are discussed. (author)
Energy Technology Data Exchange (ETDEWEB)
Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
1998-03-01
In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)
International Nuclear Information System (INIS)
Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.
1984-01-01
Conception of a reactor building requires large openings in the primary concrete shield for a postulated loss-of-coolant accident. Through these openings neutrons escape and produce dose rates in several parts of the reactor building. Some calculations using ANISN, DOT and essentially TRIPOLI-2 codes allow to compute the neutron dose rates at several places such as reactor containment operating floor and containment annulus. Some complementary shields are provided and the instrumentations are placed in area where the dose rate is lower. Comparisons are presented between measurements and calculations
International Nuclear Information System (INIS)
Borak, T.B.; Stinchcomb, T.G.
1979-01-01
A rapid system has been developed for computing charged-particle distributions generated in tissue by any neutron spectra less than 4 MeV. Oxygen and carbon recoils were derived from R-matrix theory, and hydrogen recoils were obtained from cross-section evaluation. Application to two quite different fission-neutron spectra demonstrates the flexibility of this method for providing spectral details of the different types of charged-particle recoils. Comparisons have been made between calculations and measurements of event-size distributions for a sphere of tissue 1 μm in diameter irradiated by these two neutron spectra. LET distributions have been calculated from computed charged-particle recoils and also derived from measurements using the conventional approximation that all charged particles traverse the chamber. The limitations of the approximation for these neutron spectra are discussed. (author)
International Nuclear Information System (INIS)
Robinson, G.S.
1985-08-01
The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent
Calculation of the dynamic air flow resistivity of fibre materials
DEFF Research Database (Denmark)
Tarnow, Viggo
1997-01-01
The acoustic attenuation of acoustic fiber materials is mainly determined by the dynamic resistivity to an oscillating air flow. The dynamic resistance is calculated for a model with geometry close to the geometry of real fibre material. The model constists of parallel cylinders placed randomly.......The second procedure is an extension to oscillating air flow of the Brinkman self-consistent procedure for dc flow. The procedures are valid for volume concentrations of cylinders less than 0.1. The calculations show that for the density of fibers of interest for acoustic fibre materials the simple self...
International Nuclear Information System (INIS)
Roussos, N.
1982-01-01
The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared [fr
International Nuclear Information System (INIS)
Wechsler, M.A.; Ramavarapu, R.; Daugherty, E.L.; Palmer, R.C.; Bullen, D.B.; Sommer, W.F.
1993-01-01
Calculations using the high-energy transport code LAHET have been made for the production of displacements, helium gas, and transmuted atoms for stainless steel (Fe-18 wt % Cr-10 wt % Ni) irradiated with spallation neutrons at energies of 100 to 1600 MeV. The damage energy cross section increased from about 250 to 350 b keV for increasing neutron energies from 100 to 1600 MeV with a spallation spectrum average of 281 barns-keV. For a displacement threshold energy of 33 eV, the corresponding spectrum-average displacement cross section is 3400 barns. The PKA spectrum was found to be fairly independent of the incident neutron energy, with an average damage energy of 0.25--0.30 MeV. The helium production cross section increased monotonically with increasing neutron energy, with a spectrum average of 0.32 barns. The maximum transmutation yield was observed near manganese (Z = 25), corresponding to a production cross section of about 0.2 barns. Relevance to fusion materials is discussed
International Nuclear Information System (INIS)
Laureau, Axel
2015-01-01
In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermal hydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions. As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons. A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of
Lucero, Catherine L.; Bentz, Dale P.; Hussey, Daniel S.; Jacobson, David L.; Weiss, W. Jason
Air entrainment is commonly added to concrete to help in reducing the potential for freeze thaw damage. It is hypothesized that the entrained air voids remain unsaturated or partially saturated long after the smaller pores fill with water. Small gel and capillary pores in the cement matrix fill quickly on exposure to water, but larger pores (entrapped and entrained air voids) require longer times or other methods to achieve saturation. As such, it is important to quantitatively determine the water content and degree of saturation in air entrained cementitious materials. In order to further investigate properties of cement-based mortar, a model based on Beer's Law has been developed to interpret neutron radiographs. This model is a powerful tool for analyzing images acquired from neutron radiography. A mortar with a known volume of aggregate, water to cement ratio and degree of hydration can be imaged and the degree of saturation can be estimated.
International Nuclear Information System (INIS)
Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.
1995-05-01
The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%
Nested element method in multidimensional neutron diffusion calculations
International Nuclear Information System (INIS)
Altiparmakov, D.V.
1983-01-01
A new numerical method is developed that is particularly efficient in solving the multidimensional neutron diffusion equation in geometrically complex systems. The needs for a generally applicable and fast running computer code have stimulated the inroad of a nonclassical (R-function) numerical method into the nuclear field. By using the R-functions, the geometrical components of the diffusion problem are a priori analytically implemented into the approximate solution. The class of functions, to which the approximate solution belongs, is chosen as close to the exact solution class as practically acceptable from the time consumption point of view. That implies a drastic reduction of the number of degrees of freedom, compared to the other methods. Furthermore, the reduced number of degrees of freedom enables calculation of large multidimensional problems on small computers
International Nuclear Information System (INIS)
Sarta Fuentes, Jose A.; Castiblanco, L.A.
2003-01-01
With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)
Neutron dose rate for {sup 252} Cf AT source in medical applications
Energy Technology Data Exchange (ETDEWEB)
Paredes, L.; Balcazar, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [UAM-I, 09340 Mexico D.F. (Mexico); Francois, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)
2006-07-01
The AAPM TG-43 modified protocol was used for the calculation of the neutron dose rate of {sup 252}Cf sources for two tissue substitute materials, five normal tissues and six tumours. The {sup 252}Cf AT source model was simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) neutron air kerma strength conversion factor, b) dose rate constant, c) radial dose function, d) geometry factor, e) anisotropy function and f) neutron dose rate. The calculated dose rate in water at 1 cm and 90 degrees from the source long axis, using the Watt fission spectrum, was D{sub n}(r{sub 0}, {theta}{sub 0})= 1.9160 cGy/h-{mu}g. When this value is compared with Rivard et al. calculation using MCNP4B code, 1.8730 cGy/h-{mu}g, a difference of 2.30% is obtained. The results for the reference neutron dose rate in other media show how small variations in the elemental composition between the tissues and malignant tumours, produce variations in the neutron dose rate up to 12.25%. (Author)
International Nuclear Information System (INIS)
Amorim, E.S. do; D'Oliveira, A.B.; Oliveira, E.C. de.
1980-11-01
A comparative study of the systems for neutronics calculations used in Los Alamos Scientific Laboratory (LASL) and Argonne National Laboratory (ANL) has been performed using benchmark results available in the literature, in order to analyse tghe convenience of using the respective codes MINX/NJOY and ETOE/MC 2 -2 for performing neutronics calculations in course at the Divisao de Estudos Avancados. (Author) [pt
Radiation doses from radiation sources of neutrons and photons by different computer calculation
International Nuclear Information System (INIS)
Siciliano, F.; Lippolis, G.; Bruno, S.G.
1995-12-01
In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoric background are concerned
Error reduction techniques for Monte Carlo neutron transport calculations
International Nuclear Information System (INIS)
Ju, J.H.W.
1981-01-01
Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas
Single event upsets calculated from new ENDF/B-VI proton and neutron data up to 150 MeV
International Nuclear Information System (INIS)
Chadwick, M.B.
1999-01-01
Single-event upsets (SEU) in microelectronics are calculated from newly-developed silicon nuclear reaction recoil data that extend up to 150 MeV, for incident protons and neutrons. Calculated SEU cross sections are compared with measured data
Application of Fast Neutron Activity for Analysing Element Content on the Air Particulate
International Nuclear Information System (INIS)
Elin Nuraini; Ngasifudin; Sunardi; Elisabeth
2003-01-01
The research on application of fast neutron activation analysis for analysing element content on the air particulate has been done. The research about analysis of the particulate matters contained in non industrial traffic territory of Surakarta and full industrial traffic territory of Karanganyar, had been done using Fast Neutron Activation Analysis Method. Fast Neutron Activation Analysis method is one of the element analysis method which it's basic principle causes radioactivity appearance from the samples after being irradiated by neutron. The qualitative analysis method is based on the measuring of specific energy which was radiated by radioactive's nucleus and quantitative analysis method is based on the measuring of the intensity of each peak gamma energy. The qualitative analysis results showed, some element were identified i.e : 51 V ; 200 Pb, 27 Al and 52 Cr. The result showed that Pb level is 2.21 ± 0.09x10 -1 mg/m 3 in non industrial traffic territory of Surakarta and 2.78 ± 0.11x10 -1 mg/m 3 full industrial traffic territory of Karanganyar, this value greater than threshold value according 6.0x10 -2 mg/m 3 . (author)
Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman
2018-06-01
The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.
International Nuclear Information System (INIS)
Corcuera, Roberto.
1975-12-01
The present work is a contribution to the neutronics calculational methods of fast neutron reactors. The first step is devoted to the analysis of the validity of the few-groups (of the order of 25) multigroup scheme, and of the transport-correction approximation for the treatment of the scattering anisotropy. This analysis includes both the reactor core, where the usual approximations are found to be satisfactory, and the reflector, where it turns out that the rapid variations of the neutron flux and of it's spectrum necessitate the improvement of the multigroup cross-sections' generation. Therefore, a zero-dimensional simple and accurate model for the average spectrum in the reflector is developed by the space-energy synthesis method. Finally using the Rayleigh-Ritz method, a model is developed in which the flux is spatially represented by an analytical function. This model is applied to the analysis of the sensitivity of reflector neutronics parameters to the variations of the cross sections [fr
Computational models for probabilistic neutronic calculation in TADSEA
International Nuclear Information System (INIS)
Garcia, Jesus A.R.; Curbelo, Jesus P.; Hernandez, Carlos R.G.; Oliva, Amaury M.; Lira, Carlos A.B.O.
2013-01-01
The Very High Temperature Reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed reactors, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. In previous studies, the conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made. The TADSEA is a pebble-bed ADS cooled by helium and moderated by graphite. In order to simulate the TADSEA correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and randomly located TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCNP code. The main reason is that there is a limited number of cells and surfaces to be defined. In this paper a computational tool, which allows to get a new geometrical model for fuel pebble to neutronic calculation with MCNPX, was presented. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. There are also compared several neutronic computational models for TADSEA's fuel pebbles in order to study heterogeneity effects. On the other hand the boundary effect given by the intersection between the pebble surface and the TRISO particles could be significative in the multiplicative properties. A model to study this e ect is also presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Maerker, R.E.; Williams, M.L.
1981-01-01
The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment.
International Nuclear Information System (INIS)
Maerker, R.E.; Williams, M.L.
1981-01-01
The Oak Ridge Poolside Facility (PSF), like the Pool Critical Assembly (PCA), is used for benchmark dosimetry measurements which can serve to validate the transport methods used in calculating the high-energy neutron fluences (> 0.1 MeV) in LWR pressure vessels required to estimate the neutron damage to the pressure vessels in the form of embrittlement. The PSF consists of an arrangement of two water gaps of 4 and 12 cm thickness separated by a simulated thermal shield and followed by a simulated pressure vessel wall and then a void box to represent a reactor cavity. The PSF is driven by the 30 MW ORR reactor, whereas the geometrically similar core of the PCA has a maximum power of only 10 KW. This paper reports the results of some calculated activities and compares them with published PSF measurements performed by HEDL and other laboratories on the so-called Westinghouse surveillance capsule perturbation experiment
THERMAL: A routine designed to calculate neutron thermal scattering
International Nuclear Information System (INIS)
Cullen, D.E.
1995-01-01
THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy
International Nuclear Information System (INIS)
Ivascu, M.
1983-10-01
Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations
International Nuclear Information System (INIS)
Bessis, J.
1977-01-01
The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program
Calculation of the Kernel scattering for thermal neutrons in H2O e D2O
International Nuclear Information System (INIS)
Leal, L.C.; Assis, J.T. de
1981-01-01
A computer code, using the Nelkin-and Butler models for the calculations of the Kernel scattering, was developed. Calculations of the thermal neutron flux in an homogeneous-and infinite medium with a 1 /v absorber in 30 energy groups were done and compared with experimental data. The reactors parameters calculated by the Hammer code (in the original version and with the new library generated by the authors' code) are presented. (E.G) [pt
Discrete-ordinates electron transport calculations using standard neutron transport codes
International Nuclear Information System (INIS)
Morel, J.E.
1979-01-01
The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure
International Nuclear Information System (INIS)
Matausek, M.V.; Zmijatevic, I.
1981-01-01
A procedure to solve the space-single-lethargy dependent transport equation for epithermal neutrons in a cylindricised multi-region reactor lattice cell has been developed and proposed in the earlier papers. Here, the computational algorithm is comprised and the computing program SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, as well as the related integral quantities as reaction rates and resonance integrals, is described. (author)
A comparison of the moisture gauge and the neutron log in air-filled holes at NTS
International Nuclear Information System (INIS)
Hearst, J.R.; Carlson, R.C.
1993-08-01
Two methods are commonly used to measure water content of geologic materials by neutron diffusion, the moisture gauge and the neutron log. Both are used at NTS, the moisture gauge in tunnels, the neutron log in vertical drilled holes. In this work, the moisture gauge and the neutron log are compared for use in air-filled holes NTS. The measurement instruments have evolved with very different operational characteristics and one important physics difference, the source to detector spacing. The moisture gauge has a very short, 0--6 cm spacing, with little internal shielding, and count increases with water. The neutron log has a long spacing, 30--50 cm, substantial internal shielding, and exhibits decreasing count with increasing water. The moisture gauge gives better bed resolution than the neutron log. Because its count increases with water, the moisture gauge is more strongly affected by water in the borehole, especially in dry formations. In these conditions the neutron log is the method of choice. In air-filled holes, if source size or logging time is not a constraint, the relative sensitivity of the two tools to water is determined by the relative strengths of borehole effects as fluid, holesize, or tool-wall gap. If source size is a constraint for safety reasons, the short spacing provides higher countrates for a given detector efficiency and thus better relative precision in determining the true count. If source size is limited because of detector or electronics saturation, the short spacing will be better at high water content, while the long spacing will be better at low water content. The short spacing may have an advantage because it can make better contact with the hole wall and can be more easily corrected for gap. The long spacing tool is currently used in vertical holes at NTS because that is the only tool available from logging contractors. Since they are most concerned with high water contents, the short spacing tool could prove to be better
Concise four-vector scheme for neutron transport calculations
International Nuclear Information System (INIS)
Kulacsy, K.; Lukacs, B.; Racz, A.
1995-01-01
An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs
International Nuclear Information System (INIS)
Hertel, Nolan E.; Sweezy, Jeremy; Sauber, Jeremiah S.; Vaughn, David; Cook, Andrew; Tays, Jeff; Ro, Tae-Ik
2001-01-01
dose equivalent, the instrument was used to measure the dose rate in four separate neutron fields at the NNRC: un-moderated 252 Cf, D 2 O-moderated 252 Cf, polyethylene-moderated 252 Cf, and the WEP neutron howitzer with the 252 Cf at its center. For each of these sources, dose equivalent measurements were performed at source-detector centerline distances ranging from 50 to 200 cm. The reading of the neutron probe in μSv/h was recorded at each position. The Eisenhauer, Schwartz, and Johnson (ESJ) method was used to adjust the measured dose rates for room-returned and air-scattered neutrons. A calibration factor for the instrument is then calculated by dividing the room and air-scatter corrected or 'true' dose rate at 1 m from the source by the theoretical dose at 1 m. Theoretical doses at 1 m in this work were calculated using the MCNP code. Complete modeling of the 252 Cf source geometry and materials as well as any associated source moderators was performed. Fluxes were calculated at 1 m from the sources and converted to ambient dose equivalent using dose conversion coefficients. The errors in the computed ambient dose equivalents range from 1.2% for the un-moderated 252 Cf source to 3.2% for the WEP neutron howitzer calculations. The calibration factors for the ambient dose equivalent readings obtained by dividing the air-scatter and room-return corrected dose rates by the theoretical dose rates are shown in Table I. The agreement is good and perhaps better than it should be since the Air-Scatter values used in the ESJ fit were for the count rate responses of a 25.4-cm-diam detector and not the dose rate response. More measurements with the LB 6411 neutron probe and more detailed analyses of the response data are planned. (authors)
Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.
El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H
2008-09-01
Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.
VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation
International Nuclear Information System (INIS)
Teuchert, E.; Haas, K.A.
1995-01-01
1 - Description of problem or function: VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (steady state and transient). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK. 2 - Method of solution: The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron
Comparison of neutron and high-energy X-ray dual-beam radiography for air cargo inspection
International Nuclear Information System (INIS)
Liu, Y.; Sowerby, B.D.; Tickner, J.R.
2008-01-01
Dual-beam radiography techniques utilising various combinations of high-energy X-rays and neutrons are attractive for screening bulk cargo for contraband such as narcotics and explosives. Dual-beam radiography is an important enhancement to conventional single-beam X-ray radiography systems in that it provides additional information on the composition of the object being imaged. By comparing the attenuations of transmitted dual high-energy beams, it is possible to build a 2D image, colour coded to indicate material. Only high-energy X-rays, gamma-rays and neutrons have the required penetration to screen cargo containers. This paper reviews recent developments and applications of dual-beam radiography for air cargo inspection. These developments include dual high-energy X-ray techniques as well as fast neutron and gamma-ray (or X-ray) radiography systems. High-energy X-ray systems have the advantage of generally better penetration than neutron systems, depending on the material being interrogated. However, neutron systems have the advantage of much better sensitivity to material composition compared to dual high-energy X-ray techniques. In particular, fast neutron radiography offers the potential to discriminate between various classes of organic material, unlike dual energy X-ray techniques that realistically only offer the ability to discriminate between organic and metal objects
Schrön, Martin; Köhler, Mandy; Bannehr, Lutz; Köhli, Markus; Fersch, Benjamin; Rebmann, Corinna; Mai, Juliane; Cuntz, Matthias; Kögler, Simon; Schröter, Ingmar; Wollschläger, Ute; Oswald, Sascha; Dietrich, Peter; Zacharias, Steffen
2016-04-01
Soil moisture is a key variable for environmental sciences, but its determination at various scales and depths is still an open challenge. Cosmic-ray neutron sensing has become a well accepted and unique method to monitor an effective soil water content, covering tens of hectares in area and tens of centimeters in depth. The technology is famous for its low maintanance, non-invasiveness, continous measurement, and most importantly its large footprint and penetration depth. Beeing more representative than point data, and finer resolved plus deeper penetrating than remote-sensing products, cosmic-ray neutron derived soil moisture products provide unrivaled advantage for agriculture, regional hydrologic and land surface models. The method takes advantage of omnipresent neutrons which are extraordinarily sensitive to hydrogen in soil, plants, snow and air. Unwanted hydrogen sources in the footprint can be excluded by local calibration to extract the pure soil water information. However, this procedure is not feasible for mobile measurements, where neutron detectors are mounted on a car to do catchment-scale surveys. As a solution to that problem, we suggest strategies to correct spatial neutron data with the help of available spatial data of soil type, landuse and vegetation. We further present results of mobile rover campaigns at various scales and conditions, covering small sites from 0.2 km2 to catchments of 100 km2 area, and complex terrain from agricultural fields, urban areas, forests, to snowy alpine sites. As the rover is limited to accessible roads, we further investigated the applicability of airborne measurements. First tests with a gyrocopter at 150 to 200m heights proofed the concept of airborne neutron detection for environmental sciences. Moreover, neutron transport simulations confirm an improved areal coverage during these campaigns. Mobile neutron measurements at the ground or air are a promising tool for the detection of water sources across many
International Nuclear Information System (INIS)
Woollard, J.E.; Blue, T.E.; Gupta, N.; Gahbauer, R.A.
1998-01-01
The neutron fields produced by several moderator assemblies were evaluated using both in-phantom and in-air neutron field assessment parameters. The parameters were used to determine the best moderator assembly, from among those evaluated, for use in the accelerator-based neutron source for boron neutron capture therapy. For a 10-mA proton beam current and the specified treatment parameters, a moderator assembly consisting of a BeO moderator and a Li 2 CO 3 reflector was found to be the best moderator assembly whether the comparison was based on in-phantom or in-air neutron field assessment parameters. However, the parameters were discordant regarding the moderator thickness. The in-phantom neutron field assessment parameters predict 20 cm of BeO as the best moderator thickness, whereas the in-air neutron field assessment parameters predict 25 cm of BeO as the best moderator thickness
International Nuclear Information System (INIS)
Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.
1991-01-01
Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab
International Nuclear Information System (INIS)
Shafii, M. Ali; Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny; Ariani, Menik; Yulianti, Yanti
2010-01-01
Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k eff and other parameters are investigated.
Neutronic analysis of JET external neutron monitor response
Energy Technology Data Exchange (ETDEWEB)
Snoj, Luka, E-mail: luka.snoj@ijs.si [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom (United Kingdom); Batistoni, Paola [ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden)
2016-11-01
Highlights: • We model JET tokamak containing JET remote handling system. • We investigate effect of remote handling system on external neutron monitor response. • Remote handling system correction factors are calculated. • Integral correction factors are relatively small, i.e up to 8%. - Abstract: The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated {sup 252}Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel. Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.
Energy Technology Data Exchange (ETDEWEB)
Kosunen, A
1999-08-01
In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realisation of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). To achieve the required accuracy of the dose delivered to a patient the quality of all steps in the dosimetry procedure has to be considered. This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: a)the calibration methods of plane parallel ionisation chambers used in electron beam dosimetry, (b) the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S I ?){sup water} {sub air}, in photon beams, (c) the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d) the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that up to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionisation chambers in high energy electron beams instead of calibrations in {sup 60}Co gamma beams. In photon beam dosimetry (S I ?){sup water} {sub air} can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. The use of %odd(10) requires the elimination of the electron contamination in the photon beam. By a twin ionisation chamber technique the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation
International Nuclear Information System (INIS)
Kosunen, A.
1999-08-01
In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realisation of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). To achieve the required accuracy of the dose delivered to a patient the quality of all steps in the dosimetry procedure has to be considered. This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: a)the calibration methods of plane parallel ionisation chambers used in electron beam dosimetry, (b) the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S I ?) water air , in photon beams, (c) the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d) the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that up to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionisation chambers in high energy electron beams instead of calibrations in 60 Co gamma beams. In photon beam dosimetry (S I ?) water air can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. The use of %odd(10) requires the elimination of the electron contamination in the photon beam. By a twin ionisation chamber technique the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation). To improve the accuracy
International Nuclear Information System (INIS)
Jones, D.B.
1986-01-01
EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated
International Nuclear Information System (INIS)
Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina
2012-01-01
Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.
International Nuclear Information System (INIS)
Bogacz, J.; Loskiewicz, J.; Zazula, J.M.
1991-01-01
The use of universal neutron transport codes in order to calculate the parameters of well-logging probes presents a new approach first tried in U.S.A. and UK in the eighties. This paper deals with first such an attempt in Poland. The work is based on the use of MORSE code developed in Oak Ridge National Laboratory in U.S.A.. Using CG MORSE code we calculated neutron detector response when surrounded with sandstone of porosities 19% and 38%. During the work it come out that it was necessary to investigate different methods of estimation of the neutron flux. The stochastic estimation method as used currently in the original MORSE code (next collision approximation) can not be used because of slow convergence of its variance. Using the analog type of estimation (calculation of the sum of track lengths inside detector) we obtained results of acceptable variance (∼ 20%) for source-detector spacing smaller than 40 cm. The influence of porosity on detector response is correctly described for detector positioned at 27 cm from the source. At the moment the variances are quite large. (author). 33 refs, 8 figs, 8 tabs
Monte Carlo calculation of the cross-section of single event upset induced by 14MeV neutrons
International Nuclear Information System (INIS)
Li, H.; Deng, J.Y.; Chang, D.M.
2005-01-01
High-density static random access memory may experience single event upsets (SEU) in neutron environments. We present a new method to calculate the SEU cross-section. Our method is based on explicit generation and transport of the secondary reaction products and detailed accounting for energy loss by ionization. Instead of simulating the behavior of the circuit, we use the Monte Carlo method to simulate the process of energy deposition in sensitive volumes. Thus, we do not need to know details about the circuit. We only need a reasonable guess for the size of the sensitive volumes. In the Monte Carlo simulation, the cross-section of SEU induced by 14MeV neutrons is calculated. We can see that the Monte Carlo simulation not only can provide a new method to calculate SEU cross-section, but also can give a detailed description about random process of the SEU
Model calculation of neutron reaction data for 31P in the energy range from 0.1 to 20 MeV
International Nuclear Information System (INIS)
Li Jiangting; Ge Zhigang; Sun Xiuquan
2006-01-01
The neutron data calculation of 31 P in the energy range from 0.1 to 20 MeV was carried out. The neutron optical potential parameters for 31 P in energy range from O.1 to 20 MeV were obtained, based on the fitting of the available neutron experimental data with the code APOM94. The DWUCK4 code was used to investigate the cross section for neutron direct inelastic scattering. The re-evaluated neutron data is based on the available measured data by using the UNF code. The theoretical results reproduce the experimental data well, and the results were given in ENDF/B-6 format. (authors)
PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation
Energy Technology Data Exchange (ETDEWEB)
Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo
1999-06-01
The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)
Calculation of the effective D-d neutron energy distribution incident on a cylindrical shell sample
International Nuclear Information System (INIS)
Gotoh, Hiroshi
1977-07-01
A method is proposed to calculate the effective energy distribution of neutrons incident on a cylindrical shell sample placed perpendicularly to the direction of the deuteron beam bombarding a deuterium metal target. The Monte Carlo method is used and the Fortran program is contained. (auth.)