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Sample records for thermal multigroup cross-section

  1. NDS multigroup cross section libraries

    International Nuclear Information System (INIS)

    DayDay, N.

    1981-12-01

    A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)

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

    International Nuclear Information System (INIS)

    Gowins, G.

    1984-01-01

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

  3. Multigroup cross section library; WIMS library

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2000-01-01

    The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings

  4. ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

    Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

  5. Adjustement of multigroup cross sections using fast reactor integral data

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1982-01-01

    A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

  6. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D.K.; Hendrickson, H.R. (comps.)

    1978-10-01

    These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

  7. RZ calculations for self shielded multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  8. RZ calculations for self shielded multigroup cross sections

    International Nuclear Information System (INIS)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-01-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  9. Correction of multigroup cross sections for resolved resonance interference in mixed absorbers

    International Nuclear Information System (INIS)

    Williams, M.L.

    1982-07-01

    The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed

  10. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

    2011-07-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  11. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    International Nuclear Information System (INIS)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

    2011-01-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  12. Status of multigroup cross-section data for shielding applications

    International Nuclear Information System (INIS)

    Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

    1983-01-01

    Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

  13. Nuclear data processing and multigroup cross section generation

    International Nuclear Information System (INIS)

    Kim, Jeong Do; Kil, Chung Sub

    1996-01-01

    The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)

  14. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  15. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-02-01

    Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

  16. RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Anaf, J.

    1988-02-01

    An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

  17. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    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%

  18. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  19. Kalpakkam multigroup cross section set for fast reactor applications - status and performance

    International Nuclear Information System (INIS)

    Ramanadhan, M.M.; Gopalakrishnan, M.M.

    1986-01-01

    This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs

  20. Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

    International Nuclear Information System (INIS)

    Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

    2011-01-01

    Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

  1. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

  2. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  3. Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

    1997-01-01

    Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

  4. BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

    International Nuclear Information System (INIS)

    White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.

    1996-01-01

    A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs

  5. MPI version of NJOY and its application to multigroup cross-section generation

    Energy Technology Data Exchange (ETDEWEB)

    Alpan, A.; Haghighat, A.

    1999-07-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances

  6. MPI version of NJOY and its application to multigroup cross-section generation

    International Nuclear Information System (INIS)

    Alpan, A.; Haghighat, A.

    1999-01-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures

  7. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  8. XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections

    International Nuclear Information System (INIS)

    Ganesan, S.; Jagannathan, V.; Thiyagarajan, T.K.

    2005-01-01

    1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections

  9. JSD1000: multi-group cross section sets for shielding materials

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)

  10. COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison

    International Nuclear Information System (INIS)

    Anaf, Jaime; Chalhoub, E.S.

    1990-01-01

    1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine

  11. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    International Nuclear Information System (INIS)

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

  12. 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

  13. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  14. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  15. Collapsing of multigroup cross sections in optimization problems solved by means of the maximum principle of Pontryagin

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

    A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)

  16. COMPAR: system to compare multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 AND XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1987-11-01

    A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)

  17. COMPAR: A system for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-02-01

    A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs

  18. Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung

    2009-01-01

    KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections

  19. Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices

    International Nuclear Information System (INIS)

    Balaceanu, V.; Constantin, M.

    2001-01-01

    In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with

  20. Recent validation experience with multigroup cross-section libraries and scale

    International Nuclear Information System (INIS)

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

    1995-01-01

    This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment

  1. Measurement of thermal neutron capture cross section

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  2. Optimization of multi-group cross sections for fast reactor analysis

    International Nuclear Information System (INIS)

    Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

    2013-01-01

    The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO 2 -UO 2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)

  3. Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

    International Nuclear Information System (INIS)

    Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

    1978-01-01

    Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

  4. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  5. AMPX: a modular system for multigroup cross-section generation and manipulation

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Diggs, B.R.; Webster, C.C.; Lucius, J.L.; White, J.E.; Wright, R.Q.; Westfall, R.M.

    1978-01-01

    The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures

  6. Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

    International Nuclear Information System (INIS)

    Erradi, L.; Karouani, K.

    1994-01-01

    Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)

  7. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  8. Sensitivity of LWR fuel cycle costs to uncertainties in detailed thermal cross sections

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Becker, M.; Harris, D.R.

    1979-01-01

    Cross sections averaged over the thermal energy (< 1 or 2 eV) group have been shown to have an important economic role for light-water reactors. Cost implications of thermal cross section uncertainties at the few-group level were reported earlier. When it has been determined that costs are sensitive to a specific thermal-group cross section, it becomes desirable to determine how specific energy-dependent cross sections influence fuel cycle costs. Multigroup cross-section sensitivity coefficients vary with fuel exposure. By changing the shape of a cross section displayed on a view-tube through an interactive graphics system, one can compute the change in few-group cross section using the exposure dependent sensitivity coefficients. With the changed exposure dependent few-group cross section, a new fuel cycle cost is computed by a sequence of batch depletion, core analysis, and fuel batch cost code modules. Fuel cycle costs are generally most sensitive to cross section uncertainties near the peak of the hardened Maxwellian flux

  9. Validation of multigroup neutron cross sections for the Advanced Neutron Source against the FOEHN critical experimental measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.

    1994-01-01

    The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values

  10. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

  11. Analysis of benchmark experiments for testing the IKE multigroup cross-section libraries based on ENDF/B-III and IV

    International Nuclear Information System (INIS)

    Keinert, J.; Mattes, M.

    1975-01-01

    Benchmark experiments offer the most direct method for validation of nuclear cross-section sets and calculational methods. For 16 fast and thermal critical assemblies containing uranium and/or plutonium of different compositions we compared our calculational results with measured integral quantities, such as ksub(eff), central reaction rate ratios or fast and thermal activation (dis)advantage factors. Cause of the simple calculational modelling of these assemblies the calculations proved as a good test for the IKE multigroup cross-section libraries essentially based on ENDF/B-IV. In general, our calculational results are in excellent agreement with the measured values. Only with some critical systems the basic ENDF/B-IV data proved to be insufficient in calculating ksub(eff), probably due to Pu neutron data and U 238 fast capture cross-sections. (orig.) [de

  12. MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

    2015-01-01

    A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.

  13. Review of uncertainty files and improved multigroup cross section files for FENDL

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

    The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on ''Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL'', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on (1) experimental and calculational benchmarks, (2) preparation processed libraries for FENDL/ITER, (3) specifying procedures for improving FENDL and (4) selection of activation libraries for FENDL. (author). 1 tab

  14. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  15. Microscopic cross-section measurements by thermal neutron activation

    International Nuclear Information System (INIS)

    Avila L, J.

    1987-08-01

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

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

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

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

  17. 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

  18. Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

    1993-02-01

    A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

  19. One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

    1988-11-01

    We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

  20. ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies

    International Nuclear Information System (INIS)

    2000-01-01

    A - Description of program or function: - Format: AMPX Master Interface Library format. Number of groups: Fine Group (99 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Broad Group (39 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Gamma-Ray Interaction (GRI) Library in 44-groups. Materials: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu. Origin: ENDF/B-V; LENDL-V evaluations for 12 materials. - Format: AMPX Master Interface Library format. Number of groups: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. Weighting spectrum: Maxwellian 300 K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report). Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was

  1. Thermal neutron capture cross sections of tellurium isotopes

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. Thermal neutron capture cross sections of tellurium isotopes

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  3. Thermal neutron capture cross sections of tellurium isotopes

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

  4. A computer program with graphical user interface to plot the multigroup cross sections of WIMS-D library

    International Nuclear Information System (INIS)

    Thiyagarajan, T.K.; Ganesan, S.; Jagannathan, V.; Karthikeyan, R.

    2002-01-01

    As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper

  5. Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

    International Nuclear Information System (INIS)

    Bastos, H.F.B.N.

    1979-01-01

    In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)

  6. Preparation of lumped fission product (FP) cross sections for a multigroup library

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1984-01-01

    A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

  7. MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX

    International Nuclear Information System (INIS)

    Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; Kidman, R.B.; Weisbin, C.R.; White, J.E.

    1977-01-01

    1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the

  8. REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Ganesan, S.

    1988-01-01

    1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

  9. Formats and processing of evaluated nuclear data into multigroup cross-sections

    International Nuclear Information System (INIS)

    Motta, M.

    1984-01-01

    The first part of these lectures concerns the data in nuclear files and their manipulation. The structure of the data files as divided into the resonance region (subdivided into the resolved and the unresolved regions) and the continuum region is presented. The reactions concerned are the elastic scattering; the radiative capture and the fission methods for averaging the cross sections are given. Then, the group averaging formulas and the self-shielding factors are presented in some detail. The second part concerns a presentation of nuclear data files handling and conversion. The main libraries are listed and several maintenance computer codes presented. The way the conversion among different files is handled is also presented. The listings of several BASIC programs for different cross section calculations are given. These codes are self-guided

  10. TIMS-1, Multigroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B

    International Nuclear Information System (INIS)

    Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi

    1984-01-01

    1 - Description of problem or function: TIMS-1 is a code for calculating the group constants of heavy resonant nuclei by using ENDF/ B-4 format data. This code calculates infinitely dilute cross sections and self-shielding factors as a function of composition sigma-0 temperature T and R-parameter, where R is the ratio of ato- mic number density of two different resonant nuclei. 2 - Method of solution: In the unresolved resonance region, a ladder of resonance parameters and levels is generated with Monte Carlo method. The temperature dependent cross sections are calculated with the Breit-Wigner single-level and multi-level formula. The neutron spectrum is accurately calculated by solving numerically the neutron slowing down equation using a recurrence formula for neutron slowing down source. 3 - Restrictions on the complexity of the problem: The maximum numbers of energy groups, temperatures and compositions are 60, 4 and 10 respectively

  11. FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

    International Nuclear Information System (INIS)

    Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

    1980-05-01

    Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

  12. Application of Gauss quadratures to the calculation of resolved resonance contribution in multigroup cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1989-01-01

    A program (RESQ) based on quadratures that evaluates, from ENDF/B data, the resolved resonance contribution in group-averaged cross sections (capture, fission and scattering) was developed. Single and Multilevel Breit-Wigner parameters are accepted. Constant weighting function and zero Kelvin were considered. To assure convergence, different quadrature orders may be analysed. Results are compared with other codes' reconstruction and integration methods. (author) [pt

  13. Thermal neutron absorption cross section of small samples

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Energy Technology Data Exchange (ETDEWEB)

    Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

    2015-01-15

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  15. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    International Nuclear Information System (INIS)

    Díez, C.J.; Cabellos, O.; Martínez, J.S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties

  16. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

  19. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  1. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  2. Thermal neutron cross section measurements for technetium-99

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  4. PROF-DD, Generator of Multigroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio

    2002-01-01

    1 - Description of program or function: The code system PROF-DD generates a multi-group double-differential cross section library DDX from evaluated data in ENDF/B-IV or ENDF/B-V format. The system consists of the following five modules: PROF-DDX is the main module of the system. It calculates the multigroup DDX and stores them on a master PDS file. MCFILEF generates a control file for PROF-DDX, which contains energy group and angle bin structures. SPINPTF prepares an input data file for PROF-DDX by combining the control file with other input data. DDXLIBMK edits a DDX library from the master PDS file for transport calculations. RESENDD performs resonance cross section and Doppler broadening calculations. 2 - Restrictions on the complexity of the problem: The numbers of energy groups and angle bins are less than 150 and 40, respectively

  5. THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

    International Nuclear Information System (INIS)

    Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

    1983-05-01

    The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

  6. Effective thermal neutron absorption cross section for heterogeneous mixture

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  7. Generation and performance of a multigroup coupled neutron-gamma cross-section library for deterministic and Monte Carlo borehole logging analysis

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.

    2004-01-01

    As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)

  8. FENDL/MG. Library of multigroup cross sections in GENDF and MATXS format for neutron-photon transport calculations. Version 1.1 of March 1995. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Ganesan, S.

    1996-01-01

    Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab

  9. Thermal capture cross section for 58Ni (n,γ)59 Ni reaction

    International Nuclear Information System (INIS)

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

    1989-01-01

    The 58 Ni total thermal capture cross section was determined by suming the partial cross sections calculated for the primary transitions of the reaction 58 Ni (n,γ) 59 Ni. The primary transitions energies and intensities were determined from the 58 Ni thermal neutrons prompt gamma capture gamma rays spectrum in the 3.7 to 9.3 MeV region. The obtained value for the total cross section was 4.52 + 0.10b. (author) [pt

  10. 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

  11. Re-evaluation of the neutron scattering dynamics in heavy water, generation of multigroup cross sections for THERM-126

    International Nuclear Information System (INIS)

    Keinert, J.

    1982-06-01

    In providing THERM-126 with cross section matrices for deuterium bound in heavy water the IKE phonon spectrum was reevaluated. The changes are modifications in the acoustic part and in the frequency of the second oscillator. Contrary to the phonon spectrum model for D in D 2 O in ENDF/B-IV the broad band of hindered rotations is assumed to be temperature dependent taking into account the diffusive motion of the molecule. With the new model scattering law data S (α, β) are generated in the temperature range 293.6 K-673.6 K. The THERM-126 scattering cross section matrices are calculated up to P 3 . As a validity check a lot of differential and integral cross sections are compared to experiments and benchmarks are recalculated. (orig.) [de

  12. INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: INTER calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. Version 7.01 (Jan 2005): set success flag after return from beginning. 2 - Method of solution: INTER performs integrations by using the trapezoidal rule

  13. Verification of the accuracy of Doppler broadened, self-shielded multigroup cross sections for fast power reactor applications

    International Nuclear Information System (INIS)

    Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

    1988-01-01

    Verification results for Doppler broadening and self-shielding are presented. One of the important results presented is that the original SIGMA1 method of numerical Doppler broadening has now been demonstrated to be inaccurate and not capable of producing results to within required accuracies. Fortunately, due to this study, the SIGMA1 method has been significantly improved and the new SIGMA1 is now capable of producing results to within required accuracies. Although this paper presents results based upon using only one code system, it is important to realize that the original SIGMA1 method is presently used in many cross-section processing code systems; the results of this paper indicate that unless these other code systems are updated to include the new SIGMA1 method, the results produced by these code systems could be very inaccurate. The objectives of the IAEA nuclear data processing code verification project are reviewed as well as the requirements for the accuracy of calculation of Doppler coefficients and the present status of these calculations. The initial results of Doppler broadening and self-shielding calculations are presented and the inconsistency of the results which led to the discovery of errors in the original SIGMA1 method of Doppler broadening are pointed out. Analysis of the errors found and improvements in the SIGMA1 method are presented. Improved results are presented in order to demonstrate that the new SIGMA1 method can produce results within required accuracies. Guidelines are presented to limit the uncertainty introduced due to cross-section processing in order to balance available computer resources to accuracy requirements. Finally cross-section processing code users are invited to participate in the IAEA processing code verification project in order to verify the accuracy of their calculated results. (author)

  14. Measurement of the effective thermal cross section of {sup 134}Cs by triple neutron capture reaction

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Shoji; Harada, Hideo; Katoh, Toshio [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro; Kobayashi, Katsutoshi; Motoishi, Shoji; Tanase, Masakazu

    1998-03-01

    The effective thermal cross section ({sigma}{sub eff}) of the {sup 134}Cs(n,{gamma}){sup 135}Cs reaction was measured by the activation method and the {gamma}-ray spectroscopic method in order to obtain fundamental data for research on the transmutation of nuclear wastes. The effective thermal cross section of the reaction {sup 134}Cs(n,{gamma}){sup 135}Cs was found to be 140.6{+-}8.5 barns. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  16. Bound coherent and incoherent thermal neutron scattering cross sections of the elements

    International Nuclear Information System (INIS)

    Sears, V.F.

    1982-12-01

    An up-to-date table of bound coherent and incoherent thermal neutron scattering cross sections of the elements is presented. Values from two different data sources are calculated and compared. These sources are: (1) the free-atom cross sections listed in the Σbarn bookΣ and (2) the Julich scattering length tables. We also call attention to, and clarify, the confusion that exists in the literature concerning the sign of the imaginary part of the complex scattering length

  17. Thermal-neutron fission cross section of 26. 1-min /sup 235/U/sup m/

    Energy Technology Data Exchange (ETDEWEB)

    Talbert W.L. Jr.; Starner, J.W.; Estep, R.J.; Balestrini, S.J.; Attrep M. Jr.; Efurd, D.W.; Roensch, F.R.

    1987-11-01

    The thermal-neutron fission cross section of /sup 235/U/sup m/ has been measured relative to the ground-state cross section. A rapid radiochemical separation procedure was developed to provide sizeable (10/sup 10/ to 10/sup 11/ atom) samples that were reasonably free of the parent /sup 239/Pu. From a series of eight measurements, the value of 1.42 +- 0.04 was obtained for the ratio sigma/sub m//sigma/sub g/.

  18. Thermal conduction by dark matter with velocity and momentum-dependent cross-sections

    OpenAIRE

    Vincent, Aaron C.; Scott, Pat

    2013-01-01

    We use the formalism of Gould and Raffelt to compute the dimensionless thermal conduction coefficients for scattering of dark matter particles with standard model nucleons via cross-sections that depend on the relative velocity or momentum exchanged between particles. Motivated by models invoked to reconcile various recent results in direct detection, we explicitly compute the conduction coefficients $\\alpha$ and $\\kappa$ for cross-sections that go as $v_{\\rm rel}^2$, $v_{\\rm rel}^4$, $v_{\\rm...

  19. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  20. Measurements of the effective thermal neutron absorption cross-section in multi-grain models

    International Nuclear Information System (INIS)

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

    2005-01-01

    The effective macroscopic absorption cross-section Σ a eff of thermal neutrons in a grained medium differs from the corresponding cross-section Σ a hom in the homogeneous medium consisting of the same components, contributing in the same amounts. The ratio of these cross-sections defines the grain parameter, G, which is a measure of heterogeneity of the system for neutron absorption. Heterogeneous models have been built as two- or three-component systems (Ag, Cu and Co 3 O 4 grains distributed in a regular grid in Plexiglas, in various proportions between them). The effective absorption cross-section has been measured and the experimental grain parameter has been found for each model. The obtained values are in the interval 0.34 < G < 0.58, while G = 1 means the homogeneous material. (author)

  1. Measurement of the thorium absorption cross section shape near thermal energy (LWBR development program)

    International Nuclear Information System (INIS)

    Green, L.

    1976-11-01

    The shape of the thorium absorption cross section near thermal energies was investigated. This shape is dominated by one or more negative energy resonances whose parameters are not directly known, but must be inferred from higher energy data. Since the integral quantity most conveniently describing the thermal cross section shape is the Westcottg-factor, effort was directed toward establishing this quantity to high precision. Three nearly independent g-factor estimates were obtained from measurements on a variety of foils in three different neutron spectra provided by polyethylene-moderated neutrons from a 252 Cf source and from irradiations in the National Bureau of Standards ''Standard Thermal Neutron Density.'' The weighted average of the three measurements was 0.993 +- 0.004. This is in good agreement with two recent evaluations and supports the adequacy of the current cross section descriptions

  2. Preparation of rock samples for measurement of the thermal neutron macroscopic absorption cross-section

    International Nuclear Information System (INIS)

    Czubek, J.A.; Burda, J.; Drozdowicz, K.; Igielski, A.; Kowalik, W.; Krynicka-Drozdowicz, E.; Woznicka, U.

    1986-03-01

    Preparation of rock samples for the measurement of the thermal neutron macroscopic absorption cross-section in small cylindrical two-region systems by a pulsed technique is presented. Requirements which should be fulfilled during the preparation of the samples due to physical assumptions of the method are given. A cylindrical vessel is filled with crushed rock and saturated with a medium strongly absorbing thermal neutrons. Water solutions of boric acid of well-known macroscopic absorption cross-section are used. Mass contributions of the components in the sample are specified. This is necessary for the calculation of the thermal neutron macroscopic absorption cross-section of the rock matrix. The conditions necessary for assuring the required accuracy of the measurement are given and the detailed procedure of preparation of the rock sample is described. (author)

  3. ENEA-Bologna production and testing of JEF-2.2 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V.; Blokhin, A.I.

    2005-01-01

    The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ 0 . Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)

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

    International Nuclear Information System (INIS)

    Holden, N.E.

    1994-01-01

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

  5. Thermal neutron capture cross sections resonance integrals and g-factors

    International Nuclear Information System (INIS)

    Mughabghab, S.F.

    2003-02-01

    The thermal radiative capture cross sections and resonance integrals of elements and isotopes with atomic numbers from 1 to 83 (as well as 232 Th and 238 U) have been re-evaluated by taking into consideration all known pertinent data published since 1979. This work has been undertaken as part of an IAEA co-ordinated research project on 'Prompt capture gamma-ray activation analysis'. Westcott g-factors for radiative capture cross sections at a temperature of 300K were computed by utilizing the INTER code and ENDF-B/VI (Release 8) library files. The temperature dependence of the Westcott g-factor is illustrated for 113 Cd, 124 Xe and 157 Gd at temperatures of 150, 294 and 400K. Comparisons have also been made of the newly evaluated capture cross sections of 6 Li, 7 Li, 12 C and 207 Pb with those determined by the k 0 method. (author)

  6. Assessment of the ''thermal normalization technique'' for measurement of neutron cross sections vs energy

    International Nuclear Information System (INIS)

    Peelle, R.W.; de Sassure, G.

    1977-01-01

    Refined knowledge of the thermal neutron cross sections of the fissile nuclides and of the (n,α) reaction standards, together with the reasonably well known energy dependence of the latter, have permitted resonance-region and low-keV fissile nuclide cross sections to be based on these standards together with count-rate ratios observed as a function of energy using a pulsed ''white'' source. As one evaluates cross sections for energies above 20 keV, optimum results require combination of cross section shape measurements with all available absolute measurements. The assumptions of the ''thermal normalization method'' are reviewed, and an opinion is given of the status of some of the standards required for its use. The complications which may limit the accuracy of results using the method are listed and examples are given. For the 235 U(n,f) cross section, the option is discussed of defining resonance-region fission integrals as standards. The area of the approximately 9 eV resonances in this nuclide may be known to one percent accuracy, but at present the fission integral from 0.1 to 1.0 keV is known to no better than about two percent. This uncertainty is based on the scatter among independent results, and has not been reduced by the most recent measurements. This uncertainty now limits the accuracy attainable for the 235 U(n,f) cross section below about 50 keV. Suggestions are given to indicate how future detailed work might overcome past sources of error

  7. 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)

  8. Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

    International Nuclear Information System (INIS)

    Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A.; Hentati, A.

    2012-01-01

    Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of 157 Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of nat Gd which is (49360 ± 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1σ, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 ± 500) b. (authors)

  9. Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Leconte, P.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Hentati, A. [International School in Nuclear Engineering, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

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

    International Nuclear Information System (INIS)

    Adib, M.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Tan, V H; Son, P N

    2016-01-01

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

  12. Evaluation of the neutron cross sections of 235U in the thermal energy region. Final report

    International Nuclear Information System (INIS)

    Leonard, B.R. Jr.; Kottwitz, D.A.; Thompson, J.K.

    1976-02-01

    The objective of this work has been to improve the knowledge of the thermal cross sections of the fissile nuclei as a step toward providing a standard data base for the nuclear industry. The methodology uses a form of the Adler-Adler multilevel-fission theory and Breit-Wigner multilevel-scattering theory. It incorporates these theories in a general nonlinear least-squares (LSQ) fitting program SIGLEARNThe analysis methodology in this work was applied to the thermal data on 235 U. A reference data file has been developed which includes most of the known data of interest. The first important result of this work is the assessment of the shape uncertainties of the partial cross sections. The results of our studies lead to the following values and error estimates for 235 U g factors in a thermal (20.44 0 C) energy spectrum: g/sub f/ = 0.97751 (+-0.11%); g/sub γ/ = 0.98230 (+-0.14%). A second important result of this study is the development of a recommended set of 2200 m/s (0.0253 eV) values of the parameters and the probable range of further adjustment which might be made. The analysis also provides the result of a common interpretation of energy-dependent absolute cross-section data of different measurements to yield a consistent set of experimental 0.0253 eV values with rigorous error estimates. It also provides normalization factors for relative fission and capture cross sections on a common basis with rigorous error estimates. The results of these analyses provide a basis for deciding what new measurements would be most beneficial. The most important of these would be improved direct capture data in the thermal region

  13. Thermal neutron capture and resonance integral cross sections of {sup 45}Sc

    Energy Technology Data Exchange (ETDEWEB)

    Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Thi Hien, Nguyen [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Guinyun, E-mail: gnkim@knu.ac.kr [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Kwangsoo [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shin, Sung-Gyun; Cho, Moo-Hyun [Department of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Lee, Manwoo [Research Center, Dongnam Institute of Radiological and Medical Science, Busan 619-953 (Korea, Republic of)

    2015-11-01

    The thermal neutron cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been measured relative to that of the {sup 197}Au(n,γ){sup 198}Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (G{sub th}) and resonance (G{sub epi}) neutron self-shielding, the γ-ray attenuation (F{sub g}) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been determined relative to the reference values of the {sup 197}Au(n,γ){sup 198}Au reaction, with σ{sub o,Au} = 98.65 ± 0.09 barn and I{sub o,Au} = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σ{sub o,Sc} = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be I{sub o,Sc} = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

  14. Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

    International Nuclear Information System (INIS)

    Rahman, Mafizur; Takano, Hideki

    2001-01-01

    A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

  15. Thermal response test data of five quadratic cross section precast pile heat exchangers.

    Science.gov (United States)

    Alberdi-Pagola, Maria

    2018-06-01

    This data article comprises records from five Thermal Response Tests (TRT) of quadratic cross section pile heat exchangers. Pile heat exchangers, typically referred to as energy piles, consist of traditional foundation piles with embedded heat exchanger pipes. The data presented in this article are related to the research article entitled "Comparing heat flow models for interpretation of precast quadratic pile heat exchanger thermal response tests" (Alberdi-Pagola et al., 2018) [1]. The TRT data consists of measured inlet and outlet temperatures, fluid flow and injected heat rate recorded every 10 min. The field dataset is made available to enable model verification studies.

  16. Thermal response test data of five quadratic cross section precast pile heat exchangers

    Directory of Open Access Journals (Sweden)

    Maria Alberdi-Pagola

    2018-06-01

    Full Text Available This data article comprises records from five Thermal Response Tests (TRT of quadratic cross section pile heat exchangers. Pile heat exchangers, typically referred to as energy piles, consist of traditional foundation piles with embedded heat exchanger pipes. The data presented in this article are related to the research article entitled “Comparing heat flow models for interpretation of precast quadratic pile heat exchanger thermal response tests” (Alberdi-Pagola et al., 2018 [1]. The TRT data consists of measured inlet and outlet temperatures, fluid flow and injected heat rate recorded every 10 min. The field dataset is made available to enable model verification studies.

  17. The Cross-Section for the Radiative Capture of Thermal Neutrons by Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1942-07-01

    This report is based on an experiment performed at the Cavendish Laboratory (Cambridge) by E. Broda, J. Guéron and L. Kowarski in July 1942 where the intensity of the beta-activity induced in uranium by thermal neutrons has been compared with that induced in manganese or iodine. Care was taken to avoid losses due to a Szilard-Chalmers effect. The capture cross section of uranium for thermal neutrons is found to amount to (2.78 ±0.1)*10{sup -24} cm{sup 2}, assuming the value 581*10{sup -24} cm{sup 2} for σ{sub B}. (nowak)

  18. Thermal neutron capture cross section of chromium, vanadium, titanium and nickel isotopes

    International Nuclear Information System (INIS)

    Venturini, L.; Pecequilo, B.R.S.

    1990-04-01

    The thermal neutron cross section of chromium, vanadium, titanium and nickel can be determined by measuring the pair spectrum of prompt gamma-rays emitted targets of these elements are irradiated by a thermal neutron beam. Such measurements were carried out by irradiating the natural element mixed with a nitrogen standard (melamine) in the tangential beam hole of the IEA-R1 research reactor. The pair spectrometer efficiency calibration curve in the 1.5 to 11 MeV energy range was performed with a melamine plus ammonium chloride mixed target. The cross section was calculated for the most prominent gamma transitions of each isotope, using nitrogen as standard and averaged over the obtained values. The resulting mean cross sections are as follows: (13.4 ± 0.7)b for 50 Cr, (0.79 ± 0,02)b for 52 Cr, (18.1 ± 0,7)b for 53 Cr, (4.9 ± 0.2)b for 51 V, (8.4 ± 0.1)b for 48 Ti, (4.41 ± 0.08)b 58 Ni, (2.54 ± 0.07)b for 60 Ni, (15.2 ± 0.5)b for 62 Ni and (1.6 ± 0.1) for 64 Ni. (author) [pt

  19. Thermal neutron absorption cross-section for small samples (experiments in cylindrical geometry)

    International Nuclear Information System (INIS)

    Czubek, J.A.; Drozdowicz, K.; Igielski, A.; Krynicka-Drozdowicz, E.; Woznicka, U.

    1982-01-01

    Measurement results for thermal neutron macroscopic absorption cross-sections Σsub(a)1 when applying the cylindrical sample-moderator system are presented. Experiments for liquid (water solutions of H 3 BO 3 ) and solid (crushed basalts) samples are reported. Solid samples have been saturated with the H 3 BO 3 ''poisoning'' solution. The accuracy obtained for the determination of the absorption cross-section of the solid material was σ(Σsub(ma))=(1.2+2.2) c.u. in the case when porosity was measured with the accuracy of σ(phi)=0.001+0.002. The dispersion of the Σsub(ma) data obtained for basalts (taken from different quarries) was higher than the accuracy of the measurement. All experimental data for the fundamental decay constants lambda 0 together with the whole information about the samples are given. (author)

  20. Measurement of {sup 238}Np fission cross-section by neutrons near thermal point (preliminary results)

    Energy Technology Data Exchange (ETDEWEB)

    Abramo; vich, S.N.; Andreev, M.F.; Bol`shakov, Y.M. [Institute of Experimental Physics, Arzamas (Russian Federation)] [and others

    1995-10-01

    Measurements have been carried out of {sup 238}Np fission cross-section by thermal neutrons. The isotope {sup 238}Np was built up through the reaction {sup 238}U(p,n) on an electrostatic accelerator. Extraction and cleaning of the sample were done by ion-exchange chromatography. Fast neutrons were generated on the electrostatic accelerator through the reaction {sup 9}Be(d,n); a polyethylene block was used to slow down neutrons. Registration of fission fragments was performed with dielectric track detectors. Suggesting that the behavior of {sup 238}Np and {sup 238}U. Westscott`s factors are indentical the fission cross-section of {sup 238}Np was obtained: {sigma}{sub fo}=2110 {plus_minus} 75 barn.

  1. Neutron-capture-activation cross sections of 9496Zr and 98100Mo at thermal and 30 keV energy

    International Nuclear Information System (INIS)

    Wyrick, J.M.; Poenitz, W.P.

    1982-01-01

    Neutron-capture cross sections of 94 96 Zr and 98 100 Mo were measured relative to the standard-capture cross section of gold at thermal and 30 keV neutron energies using the activation technique. The reported values are based upon available decay-scheme information

  2. Evaluation of thermal neutron cross-sections and resonance integrals of protactinium, americium, curium, and berkelium isotopes

    International Nuclear Information System (INIS)

    Belanova, T.S.

    1994-12-01

    Data on the thermal neutron fission and capture cross-sections as well as their corresponding resonance integrals are reviewed and analysed. The data are classified according to the form of neutron spectra under investigation. The weighted mean values of the cross-sections and resonance integrals for every type of neutron spectra were adopted as evaluated data. (author). 87 refs, 2 tabs

  3. 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)

  4. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  5. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

  6. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

    Science.gov (United States)

    Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

    2017-09-01

    The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  7. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2 multi-group (199 n + 42 γ cross section library in AMPX format for nuclear fission applications

    Directory of Open Access Journals (Sweden)

    Pescarini Massimo

    2017-01-01

    Full Text Available The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data. VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  8. Monte Carlo Calculation of Thermal Neutron Inelastic Scattering Cross Section Uncertainties by Sampling Perturbed Phonon Spectra

    Science.gov (United States)

    Holmes, Jesse Curtis

    Nuclear data libraries provide fundamental reaction information required by nuclear system simulation codes. The inclusion of data covariances in these libraries allows the user to assess uncertainties in system response parameters as a function of uncertainties in the nuclear data. Formats and procedures are currently established for representing covariances for various types of reaction data in ENDF libraries. This covariance data is typically generated utilizing experimental measurements and empirical models, consistent with the method of parent data production. However, ENDF File 7 thermal neutron scattering library data is, by convention, produced theoretically through fundamental scattering physics model calculations. Currently, there is no published covariance data for ENDF File 7 thermal libraries. Furthermore, no accepted methodology exists for quantifying or representing uncertainty information associated with this thermal library data. The quality of thermal neutron inelastic scattering cross section data can be of high importance in reactor analysis and criticality safety applications. These cross sections depend on the material's structure and dynamics. The double-differential scattering law, S(alpha, beta), tabulated in ENDF File 7 libraries contains this information. For crystalline solids, S(alpha, beta) is primarily a function of the material's phonon density of states (DOS). Published ENDF File 7 libraries are commonly produced by calculation and processing codes, such as the LEAPR module of NJOY, which utilize the phonon DOS as the fundamental input for inelastic scattering calculations to directly output an S(alpha, beta) matrix. To determine covariances for the S(alpha, beta) data generated by this process, information about uncertainties in the DOS is required. The phonon DOS may be viewed as a probability density function of atomic vibrational energy states that exist in a material. Probable variation in the shape of this spectrum may be

  9. 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)

  10. Revisiting the U-238 thermal capture cross section and gamma-raymission probabilities from Np-239 decay

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A.; Molnar, G.L.; Revay, Zs.; Mughabghab, S.F.; Firestone,R.B.; Pronyaev, V.G.; Nichols, A.L.; Moxon, M.C.

    2005-03-03

    The precise value of the thermal capture cross section of238U is uncertain, and evaluated cross sections from various sourcesdiffer by more than their assigned uncertainties. A number of theoriginal publications have been reviewed to assess the discrepant data,corrections were made for more recent standard cross sections andotherconstants, and one new measurement was analyzed. Due to the strongcorrelations in activation measurements, the gamma-ray emissionprobabilities from the beta decay of 239Np were also analyzed. As aresult of the analysis, a value of 2.683 +- 0.012 barns was derived forthe thermal capture cross section of 238U. A new evaluation of thegamma-ray emission probabilities from 239Np decay was alsoundertaken.

  11. ENEA-Bologna production and testing of Jeff-3.1 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Sinitsa, V.

    2008-01-01

    The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)

  12. Thermal-neutron fission cross section of 26.1-min /sup 235/U/sup m/

    International Nuclear Information System (INIS)

    Talbert, W.L. Jr.; Starner, J.W.; Estep, R.J.; Balestrini, S.J.; Attrep, M. Jr.; Efurd, D.W.; Roensch, F.R.

    1987-01-01

    The thermal-neutron fission cross section of /sup 235/U/sup m/ has been measured relative to the ground-state cross section. A rapid radiochemical separation procedure was developed to provide sizeable (10/sup 10/ to 10/sup 11/ atom) samples that were reasonably free of the parent /sup 239/Pu. From a series of eight measurements, the value of 1.42 +- 0.04 was obtained for the ratio σ/sub m//σ/sub g/

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

    International Nuclear Information System (INIS)

    Belgya, T.

    2006-01-01

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

  14. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gentry, Cole [Univ. of Tennessee, Knoxville, TN (United States); Wiarda, Dorothea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Wilmington, NC (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

  15. Measurement of thermal neutron cross section for {sup 241}Am(n,f) reaction

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Katsuhei; Yamamoto, Shuji; Fujita, Yoshiaki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Miyoshi, Mitsuharu; Kimura, Itsuro; Kanno, Ikuo; Shinohara, Nobuo

    1997-03-01

    Making use of a standard neutron spectrum field with a pure Maxwellian distribution, the thermal neutron cross section for the {sup 241}Am(n,f) reaction has been measured relative to the reference value of 586.2b for the {sup 235U}(n,f) reaction. For the present measurement, electrodeposited layers of {sup 241}Am and {sup 235}U have been employed as back-to-back type double fission chambers. The present result at neutron energy of 0.0253 eV is 3.15 {+-} 0.097b. The ENDF/B-VI data is in good agreement with the present value, while the JENDL-3.2 data is lower by 4.2%. The evaluated data in JEF-2.2 and by Mughabghab are higher by 0.9% and 1.6%, respectively than the present result. The ratios of the earlier experimental data to the present value are distributed between 0.89 and 1.02. (author)

  16. A comparative study of cross sections at few energy groups for thermal reactors fuel cells

    International Nuclear Information System (INIS)

    Claro, L.H.; Prati, A.

    1992-01-01

    A comparative study of nuclear constant calculated with LEOPARD and WIMSD-4 codes using a typical PWR cell was done. Few groups macroscopic cross section, spectral index burnup and power distribution were analyzed. (author)

  17. Cross Section Measurements for Some Elements Suited as Thermal Spectrum indicators: Cd, Sm, Gd and Lu

    Energy Technology Data Exchange (ETDEWEB)

    Sokolowski, E; Pekarek, H; Jonsson, E

    1964-05-15

    The effective cross sections of Cd, Sm, Gd and Lu have been measured by the oscillator technique in the spectrum of the central channel of the Swedish reactor R1. For Cd, Sm and Gd the 2200 m/s cross sections were deduced on the basis of Westcott's g and s factors. The values obtained were generally in agreement with other recent values obtained by integral methods, although a systematic trend indicated that the value T{sub n} - T{sub m} = 29 {+-} 10 deg C for the neutron spectrum, measured with a fast chopper, was slightly too high. A new value of T{sub n} - T{sub m} = 22.5 {+-} 3.5 deg C was deduced and new 2200 m/s cross sections were obtained by iteration. For natural Lu, the energy dependence of the cross section is not well known. Certain assumptions about the cross section function led to unreasonably high values for the 2200 m/s cross section. Complementary differential measurements of the cross sections of Cd, Sm and Gd were made with the Rl fast chopper. For Cd and Sm the 2200 m/s cross section thus obtained agreed within experimental error with those obtained from the integral measurements. For Gd, the chopper measured value was higher, confirming earlier findings and indicating that the Westcott g factor for Gd is too high. Cd: Integral meas. : {sigma}(2200) = 2,390 {+-} 45 b; Differential meas. : {sigma}(2200) = 2,445 {+-} 25 b; Sm: Integral meas. : {sigma}(2200) = 5,880 {+-} 90 b; Differential meas. : {sigma}(2200) = 5,740 {+-} 150 b; Gd: Integral meas. : {sigma}(2200) = 46,470 {+-} 550 b; Differential meas. : {sigma}(2200) = 47,900 {+-} 700 b.

  18. Cross Section Measurements for Some Elements Suited as Thermal Spectrum indicators: Cd, Sm, Gd and Lu

    Energy Technology Data Exchange (ETDEWEB)

    Sokolowski, E.; Pekarek, H.; Jonsson, E.

    1964-05-15

    The effective cross sections of Cd, Sm, Gd and Lu have been measured by the oscillator technique in the spectrum of the central channel of the Swedish reactor R1. For Cd, Sm and Gd the 2200 m/s cross sections were deduced on the basis of Westcott's g and s factors. The values obtained were generally in agreement with other recent values obtained by integral methods, although a systematic trend indicated that the value T{sub n} - T{sub m} = 29 {+-} 10 deg C for the neutron spectrum, measured with a fast chopper, was slightly too high. A new value of T{sub n} - T{sub m} = 22.5 {+-} 3.5 deg C was deduced and new 2200 m/s cross sections were obtained by iteration. For natural Lu, the energy dependence of the cross section is not well known. Certain assumptions about the cross section function led to unreasonably high values for the 2200 m/s cross section. Complementary differential measurements of the cross sections of Cd, Sm and Gd were made with the Rl fast chopper. For Cd and Sm the 2200 m/s cross section thus obtained agreed within experimental error with those obtained from the integral measurements. For Gd, the chopper measured value was higher, confirming earlier findings and indicating that the Westcott g factor for Gd is too high. Cd: Integral meas. : {sigma}(2200) = 2,390 {+-} 45 b; Differential meas. : {sigma}(2200) = 2,445 {+-} 25 b; Sm: Integral meas. : {sigma}(2200) = 5,880 {+-} 90 b; Differential meas. : {sigma}(2200) = 5,740 {+-} 150 b; Gd: Integral meas. : {sigma}(2200) = 46,470 {+-} 550 b; Differential meas. : {sigma}(2200) = 47,900 {+-} 700 b.

  19. The thermal neutron absorption cross-sections, resonance integrals and resonance parameters of silicon and its stable isotopes

    International Nuclear Information System (INIS)

    Story, J.S.

    1969-09-01

    The data available up to the end of November 1968 on the thermal neutron absorption cross-sections, resonance absorption integrals, and resonance parameters of silicon and its stable isotopes are collected and discussed. Estimates are given of the mean spacing of the energy levels of the compound nuclei near the neutron binding energy. It is concluded that the thermal neutron absorption cross-section and resonance absorption integral of natural silicon are not well established. The data on these two parameters are somewhat correlated, and three different assessments of the resonance integral are presented which differ over-all by a factor of 230. Many resonances have been detected by charged particle reactions which have not yet been observed in neutron cross-section measurements. One of these resonances of Si 2 8, at E n = 4 ± 5 keV might account for the large resonance integral which is derived, very uncertainly, from integral data. The principal source of the measured resonance integral of Si 3 0 has not yet been located. The thermal neutron absorption cross-section of Si 2 8 appears to result mainly from a negative energy resonance, possibly the resonance at E n = - 59 ± 5 keV detected by the Si 2 8 (d,p) reaction. (author)

  20. The correlations between natural elements (K, U, Th) concentrations and thermal neutron absorption cross-section value (Σa) for rock samples of Carpatia area

    International Nuclear Information System (INIS)

    Swakon, J.; Cywicka-Jakiel, T.; Drozdowicz, E.; Gabanska, B.; Loskiewicz, J.; Woznicka, U.

    1991-01-01

    The paper presents a study of correlations between concentrations of potassium, uranium and thorium and thermal neutron absorption cross section in rock samples. This knowledge of correlation should help in recognizing the expansion ways and accumulation places of the elements responsible of high thermal neutron absorption cross section in some geological environments. The correlations show the existence of connections between the thermal neutron absorption cross section value and natural radioactivity elements concentration in rocks. The results confirm the existence of correlations between natural radioactive elements concentrations (particularly thorium) and thermal neutron absorption cross - section value in some rocks. (author). 12 refs, 23 figs, 6 tabs

  1. Sensitivity analysis of U238 cross section in thermal nuclear systems

    International Nuclear Information System (INIS)

    Amorim, E.S. do; D'Oliveira, A.B.; Oliveira, E.C. de; Moura Neto, C. de.

    1980-01-01

    A sensitivity analysis system is developed for assessing the implication of uncertainties in nuclear data and related computational methods for light water power reactor. Sensitivies, at equilibrium cycle condition, are carried out for the few group macroscopic cross section of the U 238 with respect to their 35 group microscopic absorption cross section using the batch depletion code SENTEAV similar to those calculation methods used in the industry. This investigation indicates that improvements are requested on specific range of energy. These results point out the direction for worth while experimental measurements based on an analysis of costs and economic benefits. (Author) [pt

  2. {sup 41}K(n, {gamma}){sup 42}K thermal and resonance integral cross section measurements

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, F.A. Jr.; Maidana, N.L.; Vanin, V.R. [Sao Paulo Univ., SP (Brazil). Lab. do Acelerador Linear; Dias, M.S.; Koskinas, M.F. [IPEN-CNEN, Sao Paulo, SP (Brazil). Lab. de Metrolgia Nuclear; Lopez-Pino, N. [Instituto Superior de Tecnolgias y Ciencias Aplicadas (InSTEC), Habana (Cuba)

    2012-07-01

    We measured the {sup 41}K thermal neutron absorption and resonance integral cross sections after the irradiation of KNO{sub 3} samples near the core of the IEA-R1 IPEN pool-type research reactor. Bare and cadmium-covered targets were irradiated in pairs with Au-Al alloy flux-monitors. The residual activities were measured by gamma-ray spectroscopy with a HPGe detector, with special care to avoid the {sup 42}K decay {beta}{sup -} emission effects on the spectra. The gamma-ray self-absorption was corrected with the help of MCNP simulations. We applied the Westcott formalism in the average neutron flux determination and calculated the depression coefficients for thermal and epithermal neutrons due to the sample thickness with analytical approximations. We obtained 1.57(4) b and 1.02(4) b, for thermal and resonance integral cross sections, respectively, with correlation coefficient equal to 0.39.

  3. Unsteady thermal field in bars of doubly connected cross section with heat generation

    International Nuclear Information System (INIS)

    Ercoli, L.; Laura, P.A.A.; Grossi, R.O.; Sanchez Sarmiento, G.

    1985-01-01

    The title problem is approximately solved by means of a variational formulation. It is shown that in the case of complicated boundary shapes of the cross section, the technique of conformal mapping is, sometimes, quite advantageous. The analytical predictions are compared with the results obtained by means of the finite element method and, in general, good engineering agreement is shown to exist. (orig.)

  4. Comparative analysis among several cross section sets

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1983-01-01

    Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author) [pt

  5. Averaged electron collision cross sections for thermal mixtures of β-alanine conformers in the gas phase

    Science.gov (United States)

    Fujimoto, Milton M.; de Lima, Erik V. R.; Tennyson, Jonathan

    2017-10-01

    A theoretical study of elastic electron scattering by gas-phase amino acid molecule β-alanine (NH2-CH2-CH2-COOH) is presented. R-matrix calculations are performed for each of the ten lowest-lying, thermally-accessible conformers of β-alanine. Eigenphase sums, resonance features, differential and integral cross sections are computed for each conformer. The positions of the low-energy shape resonance associated with the unoccupied {π }* orbital of the -COOH group are found to vary from 2.5 to 3.3 eV and the resonance widths from 0.2 to 0.5 eV depending on the conformation. The temperature-dependent population ratios are derived, based on temperature-corrected Gibbs free energies. Averaged cross sections for thermal mixtures of the 10 conformers are presented. A comparison with previous results for the α-alanine isomer is also presented.

  6. Thermal neutron capture cross section for Fe-56(n,gamma)

    Czech Academy of Sciences Publication Activity Database

    Firestone, R. B.; Belgya, T.; Krtička, M.; Bečvář, F.; Szentmiklosi, L.; Tomandl, Ivo

    2017-01-01

    Roč. 95, č. 1 (2017), č. článku 014328. ISSN 2469-9985 R&D Projects: GA ČR GA13-07117S; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : neutron cross section * gamma gamma-coincidence data Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 3.820, year: 2016

  7. Solid-state effects on thermal-neutron cross sections and on low-energy resonances

    International Nuclear Information System (INIS)

    Harvey, J.A.; Mook, H.A.; Hill, N.W.; Shahal, O.

    1982-01-01

    The neutron total cross sections of several single crystals (Si, Cu, sapphire), several polycrystalline samples (Cu, Fe, Be, C, Bi, Ta), and a fine-powder copper sample have been measured from 0.002 to 5 eV. The Cu powder and polycrystalline Fe, Be and C data exhibit the expected abrupt changes in cross section. The cross section of the single crystal of Si is smooth with only small broad fluctuations. The data on two single Cu crystals, the sapphire crystal, cast Bi, and rolled samples of Ta and Cu have many narrow peaks approx. 10 -3 eV wide. High resolution (0.3%) transmission measurements were made on the 1.057-eV resonance in 240 Pu and the 0.433-eV resonance in 180 Ta, both at room and low temperatures to study the effects of crystal binding. Although the changes in Doppler broadening with temperature were apparent, no asymmetries due to a recoilless contribution were observed

  8. Development and testing of the VITAMIN-B7/BUGLE-B7 coupled neutron-gamma multigroup cross-section libraries

    Energy Technology Data Exchange (ETDEWEB)

    Risner, J.M.; Wiarda, D.; Miller, T.M.; Peplow, D.E.; Patton, B.W.; Dunn, M.E. [Oak Ridge National Laboratory, MS 6170, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States); Parks, B.T. [U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Mail Stop O10-B3, 11555 Rockville Pike, Rockville, MD 20852 (United States)

    2011-07-01

    The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, 'unit tests' for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190. (authors)

  9. Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries

    International Nuclear Information System (INIS)

    Risner, Joel M.; Wiarda, Dorothea; Miller, Thomas Martin; Peplow, Douglas E.; Patton, Bruce W.; Dunn, Michael E.; Parks, Benjamin T.

    2011-01-01

    The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the Evaluated Nuclear Data File (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries was accomplished using diagnostic checks in AMPX, unit tests for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in LWR shielding applications, and meet the calculational uncertainty criterion in Regulatory Guide 1.190.

  10. Comparative study of few energy group of cross sections for fuel cells of thermal reactors

    International Nuclear Information System (INIS)

    1991-08-01

    A comparative study of nuclear constants calculated with LEOPARD and WIMSD-4 codes using a typical PWR cell was done. Few groups macroscopic cross section, spectral index, burnup and power distribution were analyzed. (author) and safety concern with the transport of radioactive materials, looking for the control of eventual exposure of radiation to men, properties and environment, that is: specification of radioactive materials to be transported; choice of loaded materials; specification of requisites of loaded materials; general specification for any way of transport (earth, water and air), and responsibilities and administrative requisites. (author)

  11. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    International Nuclear Information System (INIS)

    Bjerke, M.A.; Webster, C.C.

    1981-12-01

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

  12. Pile oscillator measurements of thermal absorption cross sections of Al, Mg, Fe and Cu

    International Nuclear Information System (INIS)

    Carre, J.C.; Vidal, R.

    1964-01-01

    The phase oscillation technique used at ZOE has the property of reducing of a marked factor the effect of neutron scattering by the sample. The absorption cross sections of poorly absorbing and highly scattering materials have been measured; for neutrons of 2,200 m/s, the following values are obtained: 229 ± 3 mb for Al; 64.2 ± 1.5 mb for Mg, 2.53 ± 0.03 b for Fe and 3.74 ± 0.04 b for Cu. (authors) [fr

  13. Measurement of scattering cross sections of liquid and solid hydrogen, deuterium and deuterium hydride for thermal neutrons

    International Nuclear Information System (INIS)

    Seiffert, W.D.

    1984-01-01

    The scattering cross sections for liquid and solid normal hydrogen, para-hydrogen, deuterium and deuterium hydride were measured for thermal neutrons at various temperatures. Solid samples of para-hydrogen exhibit distinct Bragg scattering. Liquid samples of deuterium and para-hydrogen also exhibit distinct coherence phenomena, which is indicative of strong local ordering of the molecules. In para-hydrogen and deuterium hydride, the threshold for scattering with excitation of rotations is distinctly visible. The positions of the thresholds show that the molecules in liquid hydrogen are not unhindered in their movement. After the beginning of the rotational excitation the scattering cross sections of liquid and solid para-hydrogen have different shapes which is to be explained by the differences in the dynamics of the liquid and the solid specimen. 22 references

  14. Estimation of thermal neutron cross sections from K, U, Th concentrations for rock samples using neural network algorithms

    International Nuclear Information System (INIS)

    Loskiewicz, J.; Swakon, J.

    1992-01-01

    In the paper present the results of the use of the neural network algorithms to find a function Σ a =f(K, U, Th,...). The easily measurable parameters (K, U, Th concentrations, lithology) were used to estimate the thermal neutron absorption cross-section Σ a , which is difficult to measure in the borehole conditions. This paper is suggesting a possible solution to the problem. This method may have an important application in the well-logging data treatment. (author). 6 refs, 9 tabs

  15. Determination of the thermal neutron absorption cross section for rock samples by a single measurement of the time decay constant

    International Nuclear Information System (INIS)

    Krynicka, E.

    1993-01-01

    A calibration method for the determination of the thermal neutron macroscopic mass absorption cross section for rock samples is presented. The standard deviation of the final results is discussed in detail. A big advantage of the presented method is that the calibration curves have been found using the results obtained for a variety of natural rock samples of different stratigraphies and lithologies measured by Czubek's methods. An important part of the paper is a through analysis of the standard deviation of the final result. (author). 13 refs, 11 figs, 5 tabs

  16. Effective cross section values for well-moderated thermal reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Westcott, C.H

    1970-11-15

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the {sigma} (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section {sigma} being given in terms of the 2200 m/sec. value {sigma}{sub o} by the relation {sigma} {sigma}{sub o} (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  17. Effective cross section values for well-moderated thermal reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Westcott, C H

    1970-11-15

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the {sigma} (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section {sigma} being given in terms of the 2200 m/sec. value {sigma}{sub o} by the relation {sigma} {sigma}{sub o} (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  18. Effective cross section values for well-moderated thermal reactor spectra

    International Nuclear Information System (INIS)

    Westcott, C.H.

    1970-11-01

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the σ (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section σ being given in terms of the 2200 m/sec. value σ o by the relation σ σ o (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  19. Vibrational-state-selected ion--molecule reaction cross sections at thermal energies

    NARCIS (Netherlands)

    Pijkeren, D. van; Boltjes, E.; Eck, J. van; Niehaus, A.

    1984-01-01

    A method designed to measure relative ion—molecule reaction rates at thermal collision energies for selected reactant ion vibrational states is described. Relative reaction rates are determined for the three endothermic reactions: H2+ (υ)(He,H)HeH+, H2+ (υ)(Ne,H)NeH+, D2+(υ)(Ne, D)NeD+, and for the

  20. Thermal neutron absorption cross-section measured on rock samples and brines in the Institute of Nuclear Physics

    International Nuclear Information System (INIS)

    Czubek, J.A.; Drozdowicz, K.; Krynicka-Drozdowicz, E.; Igielski, A.; Woznicka, U.

    1983-01-01

    In consecutive measurements the rock sample (having a fixed and well known shape -in our case it is a sphere or a cylinder and the sample is powdered or liquid) is enveloped in shells of a plexiglass moderator (the neutron parameters of which are known) of variable thickness and irradiated with the pulsed beam of fast neutrons. The die-away rate of thermal neutrons escaping from the whole system is measured. The absorption cross-section of the sample is found as the intersection of the experimental curve (i.e. die -away rate vs thickness of the moderator) with the theoretical one. The theoretical curve is calculated for a given moderator under the assumption of a constant value of the neutron flux inside the sample. This method is independent of the value of the transport cross-section of the sample. It has been checked on artificial materials with a well known elemental composition (liquid or solid) and on the natural brines and rock samples (basalts and dolomite). A special method of calculation of the variance of the measurement has been established. It is based on the multiple computer simulations of all experimental data used in the computation. The one standard deviation of our methods is of the order of 1 up to 3 capture units (1 c.u. = 10 -3 cm -1 ). The volume of the sample needed is of the order of 500ccm. (author)

  1. FENDL multigroup libraries

    International Nuclear Information System (INIS)

    Ganesan, S.; Muir, D.W.

    1992-01-01

    Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs

  2. Using the probability method for multigroup calculations of reactor cells in a thermal energy range

    International Nuclear Information System (INIS)

    Rubin, I.E.; Pustoshilova, V.S.

    1984-01-01

    The possibility of using the transmission probability method with performance inerpolation for determining spatial-energy neutron flux distribution in cells of thermal heterogeneous reactors is considered. The results of multigroup calculations of several uranium-water plane and cylindrical cells with different fuel enrichment in a thermal energy range are given. A high accuracy of results is obtained with low computer time consumption. The use of the transmission probability method is particularly reasonable in algorithms of the programmes compiled computer with significant reserve of internal memory

  3. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  4. Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch. [CEA Saclay, Dept. d' Astrophysique de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91- Gif sur Yvette (France); Oriol, L. [CEA Cadarache, Dept. d' Etudes des Reacteurs, 13 - Saint Paul lez Durance (France); Chartier, F. [CEA Saclay, Dept. de Physico-Chimie, 91 - Gif sur Yvette (France); Mutti, P. [Institut Laue Langevin, 38 - Grenoble, (France); AlMahamid, I. [Wadsworth Center, New York State Dept. of Health, Albany, NY (United States)

    2008-07-01

    In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The {sup 232}Th, {sup 237}Np, {sup 241}Am, and {sup 244}Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)

  5. Measurements of thermal fission and capture cross sections of minor actinides within the Mini-INCA project

    International Nuclear Information System (INIS)

    Bringer, O.; Chabod, S.; Dupont, E.; Letourneau, A.; Panebianco, S.; Veyssiere, Ch.; Oriol, L.; Chartier, F.; Mutti, P.; AlMahamid, I.

    2008-01-01

    In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at Cea/DSM to determine optimal conditions for transmutation and incineration of Minor Actinides in high intensity neutron fluxes in the thermal region. Our experimental tool is based on alpha- and gamma-spectroscopy of irradiated samples and microscopic fission-chambers. It can provide both microscopic information on nuclear reactions (total and partial cross sections for neutron capture and/or fission reactions) and macroscopic information on transmutation and incineration potentials. The 232 Th, 237 Np, 241 Am, and 244 Cm transmutation chains have been explored in details, showing some discrepancies in comparison with evaluated data libraries but in overall good agreement with recent experimental data. (authors)

  6. Effect of U-238 and U-235 cross sections on nuclear characteristics of fast and thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1997-03-01

    Benchmark calculation has been made for fast and thermal reactors by using ENDF/B-VI release 2(ENDF/B-VI.2) and JENDL-3.2 nuclear data. Effective multiplication factors (k{sub eff}s) calculated for fast reactors calculated with ENDF/B-VI.2 becomes about 1% larger than the results with JENDL-3.2. The difference in k{sub eff} is caused mainly from the difference in inelastic scattering cross section of U-238. In all thermal benchmark cores, ENDF/B-VI.2 gives smaller multiplication factors than JENDL-3.2. In U-235 cores, the difference is about 0.3%dk and it becomes about 0.6% in TCA U cores. The difference in U-238 data is also important in thermal reactors, while there are found 0.1-0.3% different v values of U isotopes in thermal energy between ENDF/B-VI.2 and JENDL-3.2. (author)

  7. MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-12-01

    MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

  8. A punched-card library of neutron cross-sections and its use in the mechanized preparation of group cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations on high-speed computers

    Energy Technology Data Exchange (ETDEWEB)

    Parker, K [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

    1962-03-15

    The AWRE punched-card library of neutron cross-sections is described together with associated IBM-7090 programmes which process this data to give group-averaged cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations. The methods developed to deal with both isotropic and anisotropic elastic scattering are described. These include the multi-group transport approximation and the full treatment of anisotropic scattering using the Legendre polynomial moments of the scattering transfer matrix. The principles of group-constant formation are considered and illustrated by describing systems of group constants suitable for fast-reactor calculations. Practical problems such as the empirical adjustment of group constants to reproduce integral results and the collapsing of a many-group set of constants to give a few-group set are discussed. (author) [French] L'auteur decrit le fichier de cartes perforees sur lesquelles on enregistre a l'Atomic Weapons Research Establishment (AWRE) les sections efficaces neutroniques ainsi que les programmes IBM-7090 associes qui sont employes pour le traitement de ces informations, en vue d'obtenir des sections efficaces moyennes par groupe pouvant servir aux calculs de neutroniques a plusieurs groupes, effectues a l'aide des methodes de Monte-Carlo, S{sub n} de Carlson et autres methodes. L'auteur expose ensuite les methodes mises au point roda etudier la diffusion elastique, tant isotrope qu'anisotrope. Elles comprennent l'approximation de transport a plusieurs groupes, ainsi que le traitement complet de la diffusion anisotrope par les moments polynomiaux de Legendre de la matrice de transfert de la diffusion. L'auteur examine les principes de la formation des constantes de groupes; a titre d'illustration, il decrit les systemes de constantes de groupes qui se pretent aux calculs de reacteurs a neutrons rapides. Il expose quelques problemes pratiques, tels que l'ajustement empirique des

  9. Multiregion, multigroup collision probability method with white boundary condition for light water reactor thermalization calculations

    International Nuclear Information System (INIS)

    Ozgener, B.; Ozgener, H.A.

    2005-01-01

    A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods

  10. The total collision cross section of Ar-Ar as a function of velocity in the thermal range

    International Nuclear Information System (INIS)

    Linse, C.A.; Biesen, J.J.H. van den; Meijdenberg, C.J.N. van den

    1977-01-01

    To describe the Ar-Ar interaction several potentials have been proposed. These potentials have been derived starting from different bulk property data as well as spectroscopic and differential cross section data. The measurements of the glory structure in the total cross section as performed by Bredewout (1976) provided in principle an essential test for the existing potentials. However, the overall energy dependence of the measured cross sections was not in agreement with the theoretically predicted C 6 and C 8 values. Therefore new measurements were performed with improved angular and velocity resolution. There are still differences between the results of the measured and calculated cross sections. However, the energy dependence of the cross section remains within the limits to be expected from the theoretical predictions. (Auth.)

  11. Calculation of actual cross sections and thermalization of neutrons; Calcul des sections efficaces effectives et thermalisation des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R.

    1963-05-15

    This report gathers and presents in a simple way results of studies performed at the CEA on issues of spectra in thermal reactors. It is in fact a synthesis of results eventually published in different documents. It first presents the notion of actual cross section as it was introduced by Westcott to characterize the dependence of neutron behaviour on speed distribution. It addresses the case of a homogeneous medium with a conventional model, with the heavy gas model, and with the hydrogen gas model. It generalizes the approach by the differential model. The next part addresses the case of a heterogeneous medium, and the case of presence of moderator nuclei within the fuel [French] Le present rapport a pour objet de rassembler et de presenter de maniere simple les resultats des etudes effectuees au CE.A. sur les problemes de spectres dans les reacteurs thermiques. Ces resultats se trouvaient disperses dans plusieurs documents, ou n'etaient pas encore rediges, et bien que les etudes se poursuivent, il a paru utile d'en faire une synthese provisoire. On a cherche d'autre part a en donner une presentation elementaire, accessible aux lecteurs peu familiarises avec les problemes de thermalisation; dans cet esprit l'expose a une forme didactique, et comporte des rappels de notions bien connues comme par exemple le formalisme de Westcott. (auteur)

  12. Photoneutron cross sections measurements in 9Be, 13C e 17O with thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Semmler, Renato

    2006-01-01

    Photoneutron cross sections measurements of 9 Be, 13 C and 17 O have been obtained in the energy interval between 1,6 and 10,8 MeV, using neutron capture gamma-rays with high resolution in energy (3 a 21 eV), produced by 21 target materials, placed inside a tangential beam port, near the core of the IPEN/CNEN-SP IEA-R1 (5 MW) research reactor. The samples have been irradiated inside a 4π geometry neutron detector system 'Long Counter', 520,5 cm away from the capture target. The capture gamma-ray flux was determined by means of the analysis of the gamma spectrum obtained by using a Ge(Li) solid-state detector (EG and G ORTEC, 25 cm 3 , 5%), previously calibrated with capture gamma-rays from a standard target of Nitrogen (Melamine). The neutron photoproduction cross section has been measured for each target capture gamma-ray spectrum (compound cross section). A inversion matrix methodology to solve inversion problems for unfolding the set of experimental compound cross sections, was used in order to obtain the cross sections at specific excitation energy values (principal gamma line energies of the capture targets). The cross sections obtained at the energy values of the principal gamma lines were compared with experimental data reported by other authors, with have employed different gamma-ray sources. A good agreement was observed among the experimental data in this work with reported in the literature. (author)

  13. Neutron cross section libraries for analysis of fusion neutronics experiments

    International Nuclear Information System (INIS)

    Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo

    1988-03-01

    We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)

  14. CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

    International Nuclear Information System (INIS)

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

    1982-01-01

    Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

  15. SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 different elements plus a stainless steel

    International Nuclear Information System (INIS)

    Maerker, R.E.

    1976-01-01

    The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables

  16. Measurement of the thermal cross section of the reaction 64Zn(n,γ)65Zn in a mixed neutron flux

    International Nuclear Information System (INIS)

    Dorval, E. L; Arribere, M. A; Ribeiro Guevara, S

    2006-01-01

    Zinc is an element that is present in a great variety of biological and geological samples.For its determination by Instrumental Neutron Activation Analysis, the reaction 64 Z n(n,γ) 6 5Zn is used, due to the long half life of the reaction product, the target's bigger isotopic abundance, and an easily measurable 1115 keV main gamma line.In a recent evaluation of thermal cross sections and resonance integrals, a thermal cross section value that is 44.7% bigger than the previous evaluation has been published by the same reference.This difference is not within reported uncertainties.Besides, the relative uncertainty of the new evaluation is much bigger than the one corresponding to the previous evaluation.The adoption of the thermal cross section corresponding to different evaluations may imply, in the case of an irradiation in the peripheral I 6 position at the R A-6 reactor, discrepancies of about 43% in the calculated concentrations.These inconsistencies were evident during the irradiation of certified standard materials.This motivated the measurement of the thermal cross section for the reaction 64 Z n(n,γ) 6 5Zn in the I 6 position at the R A-6 reactor.For the analysis of results, a code was written in order to calculate correction factors from an accurate characterization of the neutron spectrum.The thermal cross section value of the reaction 64 Z n(n,γ) 6 5Zn measured is (0.76± 0.03) b [es

  17. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  18. Neutron Cross Sections for Aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Leif

    1963-08-15

    Total, elastic, inelastic, (n, 2n), (n, {alpha}), (n, p), and (n, {gamma}) cross sections for aluminium have been compiled from thermal to 100 MeV based upon literature search and theoretical interpolations and estimates. Differential elastic cross sections in the centre of mass system are represented by the Legendre coefficients. This method was chosen in order to obtain the best description of the energy dependence of the anisotropy.

  19. FEMA DFIRM Cross Sections

    Data.gov (United States)

    Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...

  20. Monte Carlo cross section testing for thermal and intermediate 235U/238U critical assemblies, ENDF/B-V vs ENDF/B-VI

    International Nuclear Information System (INIS)

    Weinman, J.P.

    1997-06-01

    The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to- 235 U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied

  1. Numerical simulation of thermal-dynamic characteristics through a helical coiled tube with annular cross section for laminar flow

    International Nuclear Information System (INIS)

    Wu Shuangying; Chen Sujun; Li Yourong; Li Longjian

    2009-01-01

    A numerical method for simulating three-dimensional laminar forced convective heat transfer in a helical coiled passage with annular cross section under uniform wall temperature condition is presented. The helical coiled passage is fabricated by bending a 0.03 m inner diameter and 0.05 m outer diameter straight tube into a helical-coil of two turns. The results presented in this paper cover a Reynolds number range of 200 ∼ 1000, a pitch range of 0.1 ∼ 0.2 and a curvature ratio range of 0.1 ∼ 0.3. The numerical computations reveal the development and distribution of heat transfer and flow fields in the helical coiled passage when the inner annular wall is heated and the outer annular wall is insulated. In addition, the effects of Reynolds number, curvature ratio, and coil pitch on the average friction factor, average Nusselt number at different axial cross-section have been discussed. The results show that the secondary flow is weak and can be neglected at the entrance region, but the effect of the secondary flow is enhanced, the maximum velocity perpendicular to axial cross section shifts toward the outer side of helical coiled passage. Furthermore, the average Nusselt number and friction factor at every different axial location present different characteristics when the Reynolds number, curvature ratio and pitch change. Compared with the curvature ratio, the pitch has relatively little influence on the heat transfer and flow performance. (authors)

  2. Burn-up physics in a coupled Hammer-Technion/Cinder-2 system and ENDF/B-V aggregate fission product thermal cross section validation

    International Nuclear Information System (INIS)

    Santos, A. dos.

    1990-01-01

    The new methodology developed in this work has the following purposes: a) to implement a burnup capability into the HAMMER-TECHNION/9/computer code by using the CINDER-2/10/computer code to perform the transmutation analysis for the actinides and fission products; b) to implement a reduced version of the CINDER-2 fission product chain structure to treat explicity nearly 99% of all original CINDER-2 fission product absorption in a typical PWR unit cell; c) to treat the effect of the fission product neutron absorption in an unit cell in a multigroup basis; d) to develop a tentative validation procedure for the ENOF/C-V stable and long-lived fission product nuclear data based on the available experimental data/11-14/. The analysis will be performed by using the reduce chain in the coupled system CINDER-2 to generate the time dependent effective four group cross sections for actinides and fission products and CINDER-2 to perform the complete transmutation analysis with its built-in chain structure. (author)

  3. Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

    Science.gov (United States)

    Pritychenko, B.; Mughabghab, S. F.

    2012-12-01

    We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

  4. Thermal neutron capture cross-section measurements of 243Am and 242Pu using the new mini-INCA α- and γ-spectroscopy station

    International Nuclear Information System (INIS)

    Marie, F.; Letourneau, A.; Fioni, G.; Deruelle, O.; Veyssiere, Ch.; Faust, H.; Mutti, P.; AlMahamid, I.; Muhammad, B.

    2006-01-01

    In the framework of the Mini-INCA project, dedicated to the study of Minor Actinide transmutation process in high neutron fluxes, an α- and γ-spectroscopy station has been developed and installed at the High Flux Reactor of the Laue-Langevin Institut. This set-up allows short irradiations as well as long irradiations in a high quasi-thermal neutron flux and post-irradiation spectroscopy analysis. It is well suited to measure precisely, in reference to 59 Co cross-section, neutron capture cross-sections, for all the actinides, in the thermal energy region. The first measurements using this set-up were done on 243 Am and 242 Pu isotopes. Cross-section values, at E n =0.025eV, were found to be (81.8+/-3.6)b for 243 Am and (22.5+/-1.1)b for 242 Pu. These values differ from evaluated data libraries by a factor of 9% and 17%, respectively, but are compatible with the most recent measurements, validating by the way the experimental apparatus

  5. Determination of the neutron resonance parameters for 206Pb and of the thermal neutron capture cross section for 206Pb and 209Bi

    International Nuclear Information System (INIS)

    Borella, A.

    2005-01-01

    response of the C6D6 detector. The analysis of the capture data allows the determination of the capture area of the resonances. In Chapter 4 we determine the thermal capture cross section for 206 Pb(n, γ) and 209 Bi(n, γ) from measurements at the cold neutron beam of the Budapest Neutron Centre. The thermal cross sections for neutron capture to the ground state 210g Bi(n, γ) and to the isomeric state 210m Bi(n, γ) have also been measured. These values complement the resonance parameters and produce a consistent description of the total and capture cross section at thermal energy and in the resolved resonance region. Chapter 5 contains the discussion of the results of this work. The statistical properties of the 206 Pb resonance parameters are described. The consistency of the resonance parameters and the thermal neutron capture cross section for 206 Pb and 209 Bi is discussed. The resulting MAC for 206 Pb is given and the impact on the termination of the s-process is described. Finally, general conclusions are presented

  6. Status of standard cross section library and future plan

    International Nuclear Information System (INIS)

    Zukeran, Atsushi

    2001-01-01

    JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group γ-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for γ-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10 -5 eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)

  7. Evaluated cross section libraries

    International Nuclear Information System (INIS)

    Maqurno, B.A.

    1976-01-01

    The dosimetry tape (ENDF/B-IV tape 412) was issued in a general CSEWG distribution, August 1974. The pointwise cross section data file was tested with specified reference spectra. A group averaged cross section data file (620 groups based on tape 412) was tested with the above spectra and the results are presented in this report

  8. Jet inclusive cross sections

    International Nuclear Information System (INIS)

    Del Duca, V.

    1992-11-01

    Minijet production in jet inclusive cross sections at hadron colliders, with large rapidity intervals between the tagged jets, is evaluated by using the BFKL pomeron. We describe the jet inclusive cross section for an arbitrary number of tagged jets, and show that it behaves like a system of coupled pomerons

  9. System THEMIS. Cross sections processing system from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented.

  10. System THEMIS. Cross sections processing system from ENDF/B

    International Nuclear Information System (INIS)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented [fr

  11. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes

    International Nuclear Information System (INIS)

    Bringer, O.

    2007-10-01

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of 241 Am and 237 Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the 241 Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  12. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

  13. Measurements of the thermal neutron cross-section and resonance integral for the 108Pd(n,γ)109Pd reaction

    Science.gov (United States)

    Hien, Nguyen Thi; Kim, Guinyun; Kim, Kwangsoo; Do, Nguyen Van; Khue, Pham Duc; Thanh, Kim Tien; Shin, Sung-Gyun; Cho, Moo-Hyun

    2018-06-01

    The thermal neutron capture cross-section (σ0) and resonance integral (I0) of the 108Pd(n,γ)109Pd reaction have been measured relative to that of the monitor reaction 197Au(n,γ)198Au. The measurements were carried out using the neutron activation with the cadmium ratio method. Both the samples and monitors were irradiated with and without cadmium cover of 0.5 mm thickness. The induced activities of the reaction products were measured with a well calibrated HPGe γ-ray detector. In order to improve the accuracy of the results, the necessary corrections for the counting losses were made. The thermal neutron capture cross-section and resonance integral of the 108Pd(n,γ)109Pd reaction were determined to be σ0,Pd = 8.68 ± 0.41 barn and I0,Pd = 245.6 ± 24.8 barn, respectively. The obtained results are compared with literature values and discussed.

  14. Floodplain Cross Section Lines

    Data.gov (United States)

    Department of Homeland Security — This table is required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally any FIRM...

  15. Multitrajectory eikonal cross sections

    International Nuclear Information System (INIS)

    Turner, R.E.

    1983-01-01

    With the use of reference and distorted transition operators, a time-correlation-function representation of the inelastic differential cross section has recently been used to obtain distorted eikonal cross sections. These cross sections involve straight-line and reference classical translational trajectories that are unaffected by any internal-state changes which have occurred during the collision. This distorted eikonal theory is now extended to include effects of internal-state changes on the translational motion. In particular, a different classical trajectory is associated with each pair of internal states. Expressions for these inelastic cross sections are obtained in terms of time-ordered cosine and sine memory functions using the Zwanzig-Feshbach projection-operator method. Explicit formulas are obtained in the time-disordered perturbation approximation

  16. Total neutron cross section of lead

    International Nuclear Information System (INIS)

    Kanda, K.; Aizawa, O.

    1976-01-01

    The total thermal-neutron cross section of natural lead under various physical conditions was measured by the transmission method. It became clear that the total cross section at room temperature previously reported is lower than the present data. The total cross section at 400, 500, and 600 0 C, above the melting point of lead, 327 0 C, was also measured, and the changes in the cross section as a function of temperature were examined, especially near and below the melting point. The data obtained for the randomly oriented polycrystalline state at room temperature were in reasonable agreement with the theoretical values calculated by the THRUSH and UNCLE-TOM codes

  17. Interference analysis of fission cross section

    International Nuclear Information System (INIS)

    Toshkov, S.A.; Yaneva, N.B.

    1976-01-01

    The formula for the reaction cross-section based on the R-matrix formalism considering the interference between the two neighbouring resonances, referred to the same value of total momentum was used for the analysis of the cross-section of resonance neutron induced fission of 230Pu. The experimental resolution and thermal motion of the target nuclei were accounted for numerical integration

  18. SENSIT: a cross-section and design sensitivity and uncertainty analysis code

    International Nuclear Information System (INIS)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE

  19. Scattering of thermal He beams by crossed atomic and molecular beams. I. Sensitivity of the elastic differential cross section to the interatomic potential

    International Nuclear Information System (INIS)

    Keil, M.; Kuppermann, A.

    1978-01-01

    The ability of diffraction oscillations in atomic beam scattering experiments to uniquely determine interatomic potentials for highly quantal systems is examined. Assumed but realistic potentials are used to generate, by scattering calculations and incorporation of random errors, differential cross sections which are then treated as if they were ''experimental'' data. From these, attempts are made to recover the initial potential by varying the parameters of assumed mathematical forms different from the original one, until a best fit to the ''experimental'' results is obtained. It is found that the region of the interaction potential around the van der Waals minimum is accurately determined by the ''measured'' differential cross sections over a range of interatomic separations significantly wider than would be expected classically. It is also found, for collision energies at which the weakly repulsive wall is appreciably sampled, that the SPF--Dunham and double Morse--van der Waals types of potentials lead to accurate determinations of the interatomic potential, whereas many other mathematical forms do not. Analytical parameterizations most appropriate for obtaining accurate interatomic potentials from thermal DCS experiments, for a given highly quantal system, may depend on the collision energy used

  20. 26Al(n,p)26Mg and 26Al(n,α)23Na cross sections from thermal energy to approximately 50 keV

    International Nuclear Information System (INIS)

    Koehler, P.E.; Gledenov, Yu.M.; Popov, Yu.P.

    1993-01-01

    Understanding the origin of 26 Al is important because it is one of the very few radioactive products of stellar nucleosynthesis to be observed directly by γ-ray telescopes. 26 Al has also been observed indirectly as a 26 Mg anomaly in some meterorites. The 26 Al(n,p) 26 Mg and 26 Al(n,α) 23 Na reactions are thought to be the major means for the destruction of 26 Al in some astrophysical environments, so a knowledge of the cross sections for these reactions is important for a better understanding of the origin of 26 Al. The authors have measured the 26 Al(n,p 1 ) 26 Mg and 26 Al(n,α 0 ) 23 Na cross sections from thermal energy to approximately 50 keV. Most of this energy range has not been explored by previous measurements. The measurements were made at the white neutron source of the Manuel Lujan, Jr. Neutron Scattering Center (LANSCE) using a ΔE-E solid-state detector telescope. Several resonances were observed. This data will be compared to previous measurements and the effect of the new data on the calculated nucleosynthesis of 26 Al will be discussed

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  2. Evaluation of cross sections of Th-232 and U-233

    International Nuclear Information System (INIS)

    Dias, A.M.

    1978-01-01

    The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

  3. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

  4. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

  5. Dependence of the Ratio between the Resonance Integral and Thermal Neutron Cross Section on the Deviation of the Epithermal Neutron Spectrum from the 1/E Law

    International Nuclear Information System (INIS)

    Soliman, N.F.

    2012-01-01

    In k 0 - Neutron Activation Analysis (k 0 -NAA), the conversion from the tabulated Q 0 (ratio of the resonance integral to thermal neutron cross-section)to Q 0 (α) (α is the shape factor of the epithermal neutron flux, indicating the deviation of the epithermal neutron spectrum from the ideal 1/E shape) are calculated using a FORTRAN program. The calculations are done for most elements that can be detected by neutron activation using different values of the parameter (α) ranging from -0.1≤α≤+0.1. The obtained data are used to study the dependence of the values (α) on the irradiation position factor in (k 0 -NAA)equation for some selected isotopes differ in their resonance energy and its Q 0 values. The results show that, the irradiation factor is affective mainly for low thermal tro epithermal flux ratio f especially for Q 0 value greater than 50. so consequently determining the irradiation parameters α value is not needed for irradiation positions that rich with thermal neutron. But for high f values the irradiation position factor should be taken into account. On the other hand the constructed FORTRAN program can be used to calculate the value Q 0 (α) directly for different value of α

  6. Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

    2010-01-01

    A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

  7. Radar cross section

    CERN Document Server

    Knott, Gene; Tuley, Michael

    2004-01-01

    This is the second edition of the first and foremost book on this subject for self-study, training, and course work. Radar cross section (RCS) is a comparison of two radar signal strengths. One is the strength of the radar beam sweeping over a target, the other is the strength of the reflected echo sensed by the receiver. This book shows how the RCS ?gauge? can be predicted for theoretical objects and how it can be measured for real targets. Predicting RCS is not easy, even for simple objects like spheres or cylinders, but this book explains the two ?exact? forms of theory so well that even a

  8. Measurement of thermal neutron cross section and resonance integral of the reaction {sup 135}Cs(n,{gamma}){sup 136}Cs

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Toshio; Nakamura, Shoji; Harada, Hideo [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan); Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro; Kobayashi, Katsutoshi; Motoishi, Shoji; Tanase, Masakazu

    1997-03-01

    The thermal neutron(2,200 m/s neutron) capture cross section({sigma}{sub 0}) and the resonance integral(I{sub 0}) of the reaction {sup 135}Cs(n,{gamma}){sup 136}Cs were measured by an activation method. Targets of radioactive cesium, which include {sup 135}Cs, {sup 137}Cs and stable {sup 133}Cs, were irradiated with reactor neutrons within or without a Cd shield case. The ratio of the number of nuclei of {sup 135}Cs to that of {sup 137}Cs was measured with a quadrupole mass spectrometer. This ratio and the ratio of activity of {sup 136}Cs to that of {sup 137}Cs were used for deduction of the {sigma}{sub 0} and the I{sub 0} of {sup 135}Cs. The {sigma}{sub 0} and the I{sub 0} of the reaction {sup 135}Cs(n,{sigma}){sup 136}Cs were 8.3 {+-} 0.3 barn and 38.1 {+-} 2.6 barn, respectively. (author)

  9. ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

    2007-01-01

    1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

  10. Generating and verification of ACE-multigroup library for MCNP

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

    2012-01-01

    The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

  11. Standard cross-section data

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1984-01-01

    The accuracy of neutron cross-section measurement is limited by the uncertainty in the standard cross-section and the errors associated with using it. Any improvement in the standard immediately improves all cross-section measurements which have been made relative to that standard. Light element, capture and fission standards are discussed. (U.K.)

  12. Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

    International Nuclear Information System (INIS)

    Lell, R.M.; Hanan, N.A.

    1987-01-01

    Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

  13. SENSIT: a cross-section and design sensitivity and uncertainty analysis code. [In FORTRAN for CDC-7600, IBM 360

    Energy Technology Data Exchange (ETDEWEB)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE.

  14. Cross sections for atmospheric corrections

    International Nuclear Information System (INIS)

    Meyer, J.P.; Casse, M.; Westergaard, N.

    1975-01-01

    A set of cross sections for spallation of relativistic nuclei is proposed based on (i) the best available proton cross sections, (ii) an extrapolation to heavier nuclei of the dependence on the number of nucleons lost of the 'target factor' observed for C 12 and O 16 by Lindstrom et al. (1975), in analogy with Rudstam's formalism, and (iii) on a normalization of all cross sections to the total cross sections for production of fragments with Asub(f) >= 6. The obtained cross sections for peripheral interactions are not inconsistent with simple geometrical considerations. (orig.) [de

  15. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  16. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  17. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2009-01-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  18. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2008-01-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  19. Covariance Evaluation Methodology for Neutron Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.

    2008-09-01

    We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.

  20. Production of a {sup 44} Ti target and its cross section of thermal neutron capture; Producao de um alvo de {sup 44} Ti e sua secao de choque para captura de neutrons termicos

    Energy Technology Data Exchange (ETDEWEB)

    Ejnisman, R

    1994-12-31

    A study of the production of a {sup 44} Ti target was carried out aiming the determination of its thermal neutron capture cross-section. With this purpose, the cross-section of the reaction {sup 45} Sc(p,2 n) {sup 44} Ti was determined in the energies 16-, 18-, 20-22- and 45 MeV. The cross-section of the reactions (p,n) {sup 45} Ti, (p,pn) {sup 44m} Sc, (p,pn) {sup 44g} Sc and (p,p2n){sup 43} Sc were also measured. The results in the low energy region are in good agreement with a previous work by McGee et al. On the other hand, the cross-section at 45 MeV is different from McGee`s result and indicates the existence of an abnormal behavior of the excitation function at higher energies. Furthermore, a radiochemical separation method was developed in order to eliminate Sc from the {sup 44} Ti target which was irradiated with neutrons. It was possible to determine an upper limit for the cross-section of the reaction {sup 44} Ti (n, {gamma}) of 4 x 10{sup 3} b. At last, it is presented a discussion of the results obtained and their possible astrophysical implications. (author) 94 refs.

  1. Neutron importance calculation in an equivalent cell using the age approximation and differential thermalization models. Determination of the cross section sensitivity to the parameters of a differential model in the thermal range

    International Nuclear Information System (INIS)

    Sidorenko, V.D.

    1978-01-01

    The equations are discussed for calculating the importance of neutron function in heterogeneous media obtained with the integral transport theory method. The thermalization effect in the thermal range is described using the differential model. The account of neutron slowing-down in the epithermal range is accomplished in the age approximation. The fast range is described in the 3-group approximation. On the basis of the equations derived the share of delayed neutrons and lifetimes of prompt neutrons are calculated and compared with available experimental data. In the thermal range the sensitivity of cross sections to some parameters of the differential model is analyzed for reactor cells typical for WWER type reactor cores. The models and approximations used are found to be adequate for the calculations

  2. Thermal neutron capture cross-section and resonance integral measurements of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce(n, γ){sup 141}Ce using a Am-Be neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Panikkath, Priyada; Mohanakrishnan, P. [Manipal University, Manipal Centre for Natural Sciences, Karnataka (India)

    2017-03-15

    Thermal neutron capture cross-sections and resonance integrals of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce (n, γ){sup 141}Ce are measured with respect to reference reactions {sup 197}Au(n, γ){sup 198}Au and {sup 55}Mn(n, γ){sup 56}Mn using the neutron activation technique. Measurements are carried out using neutrons from an Am-Be source located inside a concrete bunker. Two different methods are used for determining self-shielding factors of activation foils as well as for finding the epithermal neutron spectrum shape factor. For {sup 139}La with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 9.24 ± 0.25 b and 9.28 ± 0.37 b, respectively, while the measured resonance integrals are 12.18 ± 0.67 b and 11.81 ± 0.94 b, respectively. For {sup 140}Ce with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 0.44 ± 0.01 b and 0.44 ± 0.02 b, respectively, while the measured resonance integrals are 0.55 ± 0.03 b and 0.54 ± 0.04 b, respectively. The present measurements are compared with earlier measurements and evaluations. Presently estimated values confirm the established {sup 139}La(n, γ){sup 140}La cross-sections. The presently measured thermal capture cross-section {sup 140}Ce(n, γ){sup 141}Ce, though lower than the evaluated data, is having higher accuracy compared to previous measurements with large uncertainties. The resonance integral measured is higher (like most previous measurements) than most evaluations requiring a revision of the evaluated data. (orig.)

  3. ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE

    International Nuclear Information System (INIS)

    1970-01-01

    A - Nature of physical problem solved: Format: ANISN, DTF-4, DOT and MORSE. Number of groups: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV) Nuclides: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections). Origin: ENDF/B for H, C, N, O, Na, and Mg; O5R library for Ca, S, and K; GAM-2 library for Cl; Evaluation by J.J. Ritts for P. Weighting spectrum: 1/E for the top 99 groups and Maxwellian for the thermal group values. DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3. DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows - 1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements. 2. A 100-group set of neutron cross sections for the 11 elements. 3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle. B - Method of solution: The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values. The gamma-ray cross sections were computed from DLC-3/HPIC using MUG. The

  4. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  5. Validation of the WIMSD4M cross-section generation code with benchmark results

    International Nuclear Information System (INIS)

    Deen, J.R.; Woodruff, W.L.; Leal, L.E.

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

  6. Validation of the WIMSD4M cross-section generation code with benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

  7. Mechanized evaluation of neutron cross-sections

    International Nuclear Information System (INIS)

    Horsley, A.; Parker, J.B.

    1967-01-01

    The evaluation work to provide accurate and consistent neutron cross-section data for multigroup neutronics calculations is not fully exploiting the available theoretical and experimental results; this has been so particularly since the introduction of on-line data handling techniques enabled experimenters to turn out vast quantities of numbers. This situation can be radically improved only by mechanizing the evaluation processes. Systems such as the SC1SRS tape will not only largely overcome the task of collecting data but will provide speedy access to it; by using computers and graph-plotting machines to tabulate and display this data, the labour of evaluation can be very greatly reduced. With some types of cross-section there is hope that by using modern curve-fitting techniques the actual evaluation and statistical accounting of the data can be performed automatically. Some areas where automatic evaluation would seem likely to succeed are specified and a discussion of the mathematical difficulties incurred, such as the elimination of anomalous data, is given. Particularly promising is the use of splines in the mechanized evaluation of data. Splines are the mathematical analogues of the draughtsman's spline used in drawing smooth curves. Their principal properties are the excellent approximations they give to the derivatives of a function; in contrast to conventional polynomial fitting, this feature ensures good interpolation and, when required, stable extrapolation. Various methods of using splines in data graduation and the problem of marrying these methods to standard statistical procedures are examined. The results of work done at AWRE with cubic splines on the mechanized evaluation of neutron scattering total cross-section and angular distribution data are presented. (author)

  8. Nuclear characteristics of Pu fueled LWR and cross section sensitivities

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering

    1998-03-01

    The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)

  9. Distribution of two-phase flow thermal and hydraulic parameters over the cross-section of channels with a rod bundle

    International Nuclear Information System (INIS)

    Mironov, Yu.V.; Shpanskij, S.V.

    1975-01-01

    The paper describes PUCHOK-2, a program for thermohydraulic calculation of a channel with a bundle of smooth fuel elements. The pro.gram takes into consideration the non-uniformity of flow parameter distributions over the channel cross-section. The channel cross-section was divided into elementary cells, within which changes in flow parameters (mass velocity, heat- and steam content) were disregarded. The bundle was considered to be a system of parallel interconnected channels. Accounting for equal pressure drops in all the cells, the above model led to a system of non-linear algebraic equations. The system of equations was solved by the method of successive approximations. Theoretical results were compared with experimental data

  10. XCOM: Photon Cross Sections Database

    Science.gov (United States)

    SRD 8 XCOM: Photon Cross Sections Database (Web, free access)   A web database is provided which can be used to calculate photon cross sections for scattering, photoelectric absorption and pair production, as well as total attenuation coefficients, for any element, compound or mixture (Z <= 100) at energies from 1 keV to 100 GeV.

  11. Doppler broadening of cross sections

    International Nuclear Information System (INIS)

    Buckler, P.A.C.; Pull, I.C.

    1962-12-01

    Expressions for temperature dependent cross-sections in terms of resonance parameters are obtained, involving generalisations of the conventional Doppler functions, ψ and φ. Descriptions of Fortran sub-routines, which calculate broadened cross-sections in accordance with the derived formulae, are included. (author)

  12. ZZ THERMOS, Multigroup P0 to P5 Thermal Scattering Kernels from ENDF/B Scattering Law Data

    International Nuclear Information System (INIS)

    McCrosson, F.J.; Finch, D.R.

    1975-01-01

    1 - Description of problem or function: Number of groups: 30-group THERMOS thermal scattering kernels. Nuclides: Molecular H 2 O, Molecular D 2 O, Graphite, Polyethylene, Benzene, Zr bound in ZrHx, H bound in ZrHx, Beryllium-9, Beryllium Oxide, Uranium Dioxide. Origin: ENDF/B library. Weighting Spectrum: yes. These data are 30-group THERMOS thermal scattering kernels for P0 to P5 Legendre orders for every temperature of every material from s(alpha,beta) data stored in the ENDF/B library. These scattering kernels were generated using the FLANGE2 computer code (NESC Abstract 368). To test the kernels, the integral properties of each set of kernels were determined by a precision integration of the diffusion length equation and compared to experimental measurements of these properties. In general, the agreement was very good. Details of the methods used and results obtained are contained in the reference. The scattering kernels are organized into a two volume magnetic tape library from which they may be retrieved easily for use in any 30-group THERMOS library. The contents of the tapes are as follows - (Material: ZA/Temperatures (degrees K)): Molecular H 2 O: 100.0/296, 350, 400, 450, 500, 600, Molecular D 2 O: 101.0/296, 350, 400, 450, 500, 600, Graphite: 6000.0/296, 400, 500, 600, 700, 800, Polyethylene: 205.0/296, 350 Benzene: 106.0/296, 350, 400, 450, 500, 600, Zr bound in ZrHx: 203.0/296, 400, 500, 600, 700, 800, H bound in ZrHx: 230.0/296, 400, 500, 600, 700, 800, Beryllium-9: 4009.0/296, 400, 500, 600, 700, 800, Beryllium Oxide: 200.0/296, 400, 500, 600, 700, 800, Uranium Dioxide: 207.0/296, 400, 500, 600, 700, 800 2 - Method of solution: Kernel generation is performed by direct integration of the thermal scattering law data to obtain the differential scattering cross sections for each Legendre order. The integral parameter calculation is done by precision integration of the diffusion length equation for several moderator absorption cross sections followed by a

  13. New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

    International Nuclear Information System (INIS)

    Brown, Forrest B.

    2016-01-01

    Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.

  14. Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Shim, Hyung Jin

    2015-01-01

    The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

  15. Precise measurements of neutron capture cross sections for FP

    International Nuclear Information System (INIS)

    Nakamura, Shoji; Harada, Hideo; Katoh, Toshio

    2000-01-01

    The thermal neutron capture cross sections (σ 0 ) and the resonance integrals (I 0 ) of some fission products (FP), such as 137 Cs, 90 Sr, 99 Tc, 129 I and 135 Cs, were measured by the activation and γ-ray spectroscopic methods. Moreover, the cross section measurements were done for other FP elements, such as 127 I, 133 Cs and 134 Cs. This paper provides the summary of the FP cross section measurements, which have been performed by authors. (author)

  16. Thermal neutron calibration of a tritium extraction facility using the 6Li(n,t)4He/197Au(n,γ)198Au cross section ratio for standardization

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Smith, D.L.

    1980-08-01

    Absolute tritium activities in a neutron-activated metallic lithium samples have been measured by liquid scintillation methods to provide data needed for the determination of capture-to-fission ratios in fast breeder reactor spectra and for recent measurements of the 7 Li(n,n't) 4 He cross section. The tritium extraction facility used for all these experiments has now been calibrated by measuring the 6 Li(n,t) 4 He/ 197 Au/n,γ) 198 Au activity ratio for thermal neutrons and comparing the result with the well-known cross sections. The calculated-to-measured activity ratio was found to be 1.033 +- 0.018. 2 figures, 20 tables

  17. Photon-splitting cross sections

    International Nuclear Information System (INIS)

    Johannessen, A.M.; Mork, K.J.; Overbo, I.

    1980-01-01

    The differential cross section for photon splitting (scattering of one photon into two photons) in a Coulomb field, obtained earlier by Shima, has been integrated numerically to yield various differential cross sections. Energy spectra differential with respect to the energy of one of the outgoing photons are presented for several values of the primary photon energy. Selected examples of recoil momentum distributions and some interesting doubly or multiply differential cross sections are also given. Values for the total cross section are obtained essentially for all energies. The screening effect caused by atomic electrons is also taken into account, and is found to be important for high energies, as in e + e - pair production. Comparisons with various approximate results obtained by previous authors mostly show fair agreement. We also discuss the possibilities for experimental detection and find the most promising candidate to be a measurement of both photons, and their energies, at a moderately high energy

  18. Total Cross-Sections of U, UO{sub 2} and ThO{sub 2} for Thermal and Subthermal Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Beshai, S F [IAEA-fellow from Atomic Energy Establishment (Egypt)

    1966-03-15

    The total neutron cross-sections of U, UO{sub 2} and ThO{sub 2} have been measured from 0.0045 eV to 0.028 eV, using the time-of-flight technique. The samples were measured at 20 deg C. ThO{sub 2} was also measured at 750 deg C. The neutron source was the reactor Rl, Stockholm. The experimental results presented as graphs in the report show in detail the influence of Bragg scattering. The results further show that the increase of the temperature for the ThO{sub 2} sample gives an increase in the cross-section. The work also contains some calculations of the position in energy of Bragg edges for the three materials. These calculations show a very good agreement with the experiments. For uranium metal some calculations have been carried out also for the height ({sigma}{sup g}{sub hkl}) of the edges The agreement with the experiments is reasonable.

  19. Cross section recondensation method via generalized energy condensation theory

    International Nuclear Information System (INIS)

    Douglass, Steven; Rahnema, Farzad

    2011-01-01

    Highlights: → A new method is presented which corrects for core environment error from specular boundaries at the lattice cell level. → Solution obtained with generalized energy condensation provides improved approximation to the core level fine-group flux. → Iterative recondensation of the cross sections and unfolding of the flux provides on-the-fly updating of the core cross sections. → Precomputation of energy integrals and fine-group cross sections allows for easy implementation and efficient solution. → Method has been implemented in 1D and shown to correct the environment error, particularly in strongly heterogeneous cores. - Abstract: The standard multigroup method used in whole-core reactor analysis relies on energy condensed (coarse-group) cross sections generated from single lattice cell calculations, typically with specular reflective boundary conditions. Because these boundary conditions are an approximation and not representative of the core environment for that lattice, an error is introduced in the core solution (both eigenvalue and flux). As current and next generation reactors trend toward increasing assembly and core heterogeneity, this error becomes more significant. The method presented here corrects for this error by generating updated coarse-group cross sections on-the-fly within whole-core reactor calculations without resorting to additional cell calculations. In this paper, the fine-group core flux is unfolded by making use of the recently published Generalized Condensation Theory and the cross sections are recondensed at the whole-core level. By iteratively performing this recondensation, an improved core solution is found in which the core-environment has been fully taken into account. This recondensation method is both easy to implement and computationally very efficient because it requires precomputation and storage of only the energy integrals and fine-group cross sections. In this work, the theoretical basis and development

  20. Accurate Cross Sections for Microanalysis

    OpenAIRE

    Rez, Peter

    2002-01-01

    To calculate the intensity of x-ray emission in electron beam microanalysis requires a knowledge of the energy distribution of the electrons in the solid, the energy variation of the ionization cross section of the relevant subshell, the fraction of ionizations events producing x rays of interest and the absorption coefficient of the x rays on the path to the detector. The theoretical predictions and experimental data available for ionization cross sections are limited mainly to K shells of a...

  1. Detailed north-south cross section showing environments of deposition, organic richness, and thermal maturities of lower Tertiary rocks in the Uinta Basin, Utah

    Science.gov (United States)

    Johnson, Ronald C.

    2014-01-01

    The Uinta Basin of northeast Utah has produced large amounts of hydrocarbons from lower Tertiary strata since the 1960s. Recent advances in drilling technologies, in particular the development of efficient methods to drill and hydraulically fracture horizontal wells, has spurred renewed interest in producing hydrocarbons from unconventional low-permeability dolomite and shale reservoirs in the lacustrine, Eocene Green River Formation. The Eocene Green River Formation was deposited in Lake Uinta, a long-lived saline lake that occupied the Uinta Basin, the Piceance Basin to the east, and the intervening Douglas Creek arch. The focus of recent drilling activity has been the informal Uteland Butte member of the Green River Formation and to a much lesser extent the overlying R-0 oil shale zone of the Green River Formation. Initial production rates ranging from 500 to 1,500 barrels of oil equivalent per day have been reported from the Uteland Butte member from horizontal well logs that are as long as 4,000 feet (ft);. The cross section presented here extends northward from outcrop on the southern margin of the basin into the basin’s deep trough, located just south of the Uinta Mountains, and transects the area where this unconventional oil play is developing. The Monument Butte field, which is one of the fields located along this line of section, has produced hydrocarbons from conventional sandstone reservoirs in the lower part of the Green River Formation and underlying Wasatch Formation since 1981. A major fluvial-deltaic system entered Lake Uinta from the south, and this new line of section is ideal for studying the effect of the sediments delivered by this drainage on hydrocarbon reservoirs in the Green River Formation. The cross section also transects the Greater Altamont-Bluebell field in the deepest part of the basin, where hydrocarbons have been produced from fractured, highly overpressured marginal lacustrine and fluvial reservoirs in the Green River, Wasatch

  2. MCNP and MATXS cross section libraries based on JENDL-3.3

    International Nuclear Information System (INIS)

    Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

    2003-01-01

    The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

  3. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  4. Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

    International Nuclear Information System (INIS)

    Calloo, A.; Vidal, J. F.; Le Tellier, R.; Rimpault, G.

    2012-01-01

    This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, which better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)

  5. Criticality benchmark comparisons leading to cross-section upgrades

    International Nuclear Information System (INIS)

    Alesso, H.P.; Annese, C.E.; Heinrichs, D.P.; Lloyd, W.R.; Lent, E.M.

    1993-01-01

    For several years criticality benchmark calculations with COG. COG is a point-wise Monte Carlo code developed at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The principle consideration in developing COG was that the resulting calculation would be as accurate as the point-wise cross-sectional data, since no physics computational approximations were used. The objective of this paper is to report on COG results for criticality benchmark experiments in concert with MCNP comparisons which are resulting in corrections an upgrades to the point-wise ENDL cross-section data libraries. Benchmarking discrepancies reported here indicated difficulties in the Evaluated Nuclear Data Livermore (ENDL) cross-sections for U-238 at thermal neutron energy levels. This led to a re-evaluation and selection of the appropriate cross-section values from several cross-section sets available (ENDL, ENDF/B-V). Further cross-section upgrades anticipated

  6. Low Energy Neutrino Cross Sections

    International Nuclear Information System (INIS)

    Zeller, G.P.

    2004-01-01

    Present atmospheric and accelerator based neutrino oscillation experiments operate at low neutrino energies (Ev ∼ 1 GeV) to access the relevant regions of oscillation parameter space. As such, they require precise knowledge of the cross sections for neutrino-nucleon interactions in the sub-to-few GeV range. At these energies, neutrinos predominantly interact via quasi-elastic (QE) or single pion production processes, which historically have not been as well studied as the deep inelastic scattering reactions that dominate at higher energies.Data on low energy neutrino cross sections come mainly from bubble chamber, spark chamber, and emulsion experiments that collected their data decades ago. Despite relatively poor statistics and large neutrino flux uncertainties, these measurements provide an important and necessary constraint on Monte Carlo models in present use. The following sections discuss the current status of QE, resonant single pion, coherent pion, and single kaon production cross section measurements at low energy

  7. SCAMPI: A code package for cross-section processing

    International Nuclear Information System (INIS)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-01-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

  8. Nuclear cross section library for oil well logging analysis

    International Nuclear Information System (INIS)

    Kodeli, I.; Kitsos, S.; Aldama, D.L.; Zefran, B.

    2003-01-01

    As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) Project of the EU Community's 5 th Programme a special purpose multigroup cross section library to be used in the deterministic (as well as Monte Carlo) oil well logging particle transport calculations was prepared. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (author)

  9. SCAMPI: A code package for cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-04-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

  10. ENDF/B-5 fission product cross section evaluations

    International Nuclear Information System (INIS)

    Schenter, R.E.; England, T.R.

    1979-12-01

    Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

  11. Annotated bibliography covering generation and use of evaluated cross section uncertainty files

    International Nuclear Information System (INIS)

    Peelle, R.W.; Burrows, T.W.

    1983-03-01

    Literature references related to definition, generation, and use of evaluated cross section uncertainty (variance-covariance) files are listed with comments intended primarily to guide the reader toward materials of immediate interest. Papers are also cited that cover covariance information for individual experiments and that relate to production and use of multigroup covariance matrices. Titles are divided among several major categories

  12. Terahertz radar cross section measurements

    DEFF Research Database (Denmark)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-01-01

    We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar...

  13. Pion-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.

    1990-01-01

    The tables of inelastic and total cross sections of π ± mesons interactions with nuclei 4 He- 238 U are presented. The tables are obtained by theoretical analysis of known experimental data for energies higher some tens of MeV. 1 ref.; 1 tab

  14. Neutron capture cross section standards for BNL 325, Fourth Edition

    International Nuclear Information System (INIS)

    Holden, N.E.

    1981-01-01

    This report evaluates the experimental data and recommends values for the thermal neutron cross sections and resonance integrals for the neutron capture reactions: 55 Mn(n,γ), 59 Co(n,γ) and 197 Au(n,γ). The failure of lithium and boron as standards due to the natural variation of the absorption cross sections of these elements is discussed. The Westcott convention, which describes the neutron spectrum as a thermal Maxwellian distribution with an epithermal component, is also discussed

  15. Shielding Factor Method for producing effective cross sections: MINX/SPHINX and the CCCC interface system

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Weisbin, C.R.; Paik, N.C.

    1978-01-01

    The Shielding Factor Method (SFM) is an economical designer-oriented method for producing the coarse-group space and energy self-shielded cross sections needed for reactor-core analysis. Extensive experience with the ETOX/1DX and ENDRUN/TDOWN systems has made the SFM the method of choice for most US fast-reactor design activities. The MINX/SPHINX system was designed to expand upon the capabilities of the older SFM codes and to incorporate the new standard interfaces for fast-reactor cross sections specified by the Committee for Computer Code Coordination (CCCC). MINX is the cross-section processor. It generates multigroup cross sections, shielding factors, and group-to-group transfer matriccs from ENDF/B-IV and writes them out as CCCC ISOTXS and BRKOXS files. It features detailed pointwise resonance reconstruction, accurate Doppler broadening, and an efficient treatment of anisotropic scattering. SPHINX is the space-and-energy shielding code. It uses specific mixture and geometry information together with equivalence principles to construct shielded macroscopic multigroup cross sections in as many as 240 groups. It then makes a flux calculation by diffusion or transport methods and collapses to an appropriate set of cell-averaged coarse-group effective cross sections. The integration of MINX and SPHINX with the CCCC interface system provides an efficient, accurate, and convenient system for producing effective cross sections for use in fast-reactor problems. The system has also proved useful in shielding and CTR applications. 3 figures, 4 tables

  16. Measurement of actinide neutron cross sections

    International Nuclear Information System (INIS)

    Firestone, Richard B.; Nitsche, Heino; Leung, Ka-Ngo; Perry, DaleL.; English, Gerald

    2003-01-01

    The maintenance of strong scientific expertise is critical to the U.S. nuclear attribution community. It is particularly important to train students in actinide chemistry and physics. Neutron cross-section data are vital components to strategies for detecting explosives and fissile materials, and these measurements require expertise in chemical separations, actinide target preparation, nuclear spectroscopy, and analytical chemistry. At the University of California, Berkeley and the Lawrence Berkeley National Laboratory we have trained students in actinide chemistry for many years. LBNL is a leader in nuclear data and has published the Table of Isotopes for over 60 years. Recently, LBNL led an international collaboration to measure thermal neutron capture radiative cross sections and prepared the Evaluated Gamma-ray Activation File (EGAF) in collaboration with the IAEA. This file of 35, 000 prompt and delayed gamma ray cross-sections for all elements from Z=1-92 is essential for the neutron interrogation of nuclear materials. LBNL has also developed new, high flux neutron generators and recently opened a 1010 n/s D+D neutron generator experimental facility

  17. Neutron cross sections for fusion

    International Nuclear Information System (INIS)

    Haight, R.C.

    1979-10-01

    First generation fusion reactors will most likely be based on the 3 H(d,n) 4 He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references

  18. Negative ion detachment cross sections

    International Nuclear Information System (INIS)

    Champion, R.L.; Doverspike, L.D.

    1992-10-01

    The authors have measured absolute cross sections for electron detachment and charge exchange for collision of O and S with atomic hydrogen, have investigated the sputtering and photodesorption of negative ions from gas covered surfaces, and have begun an investigation of photon-induced field emission of electrons from exotic structures. Brief descriptions of these activities as well as future plans for these projects are given below

  19. Fission cross section measurements of actinides at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Tovesson, Fredrik [Los Alamos National Laboratory; Laptev, Alexander B [Los Alamos National Laboratory; Hill, Tony S [INL

    2010-01-01

    Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications. By combining measurement at two LANSCE facilities, Lujan Center and the Weapons Neutron Research center (WNR), differential cross sections can be measured from sub-thermal energies up to 200 MeV. Incident neutron energies are determined using the time-of-flight method, and parallel-plate ionization chambers are used to measure fission cross sections relative to the {sup 235}U standard. Recent measurements include the {sup 233,238}U, {sup 239,242}Pu and {sup 243}Am neutron-induced fission cross sections. In this paper preliminary results for cross section data of {sup 243}Am and {sup 233}U will be presented.

  20. Microscopic cross sections: An utopia?

    Energy Technology Data Exchange (ETDEWEB)

    Hilaire, S. [CEA Bruyeres-le-Chatel, DIF 91 (France); Koning, A.J. [Nuclear Research and Consultancy Group, PO Box 25, 1755 ZG Petten (Netherlands); Goriely, S. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Campus de la Plaine, CP 226, 1050 Brussels (Belgium)

    2010-07-01

    The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations. While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

  1. Microscopic cross sections: An utopia?

    International Nuclear Information System (INIS)

    Hilaire, S.; Koning, A.J.; Goriely, S.

    2010-01-01

    The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations.While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

  2. ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

    Science.gov (United States)

    Brown, D. A.; Chadwick, M. B.; Capote, R.; Kahler, A. C.; Trkov, A.; Herman, M. W.; Sonzogni, A. A.; Danon, Y.; Carlson, A. D.; Dunn, M.; Smith, D. L.; Hale, G. M.; Arbanas, G.; Arcilla, R.; Bates, C. R.; Beck, B.; Becker, B.; Brown, F.; Casperson, R. J.; Conlin, J.; Cullen, D. E.; Descalle, M.-A.; Firestone, R.; Gaines, T.; Guber, K. H.; Hawari, A. I.; Holmes, J.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Koning, A. J.; Kopecky, S.; Leal, L.; Lestone, J. P.; Lubitz, C.; Márquez Damián, J. I.; Mattoon, C. M.; McCutchan, E. A.; Mughabghab, S.; Navratil, P.; Neudecker, D.; Nobre, G. P. A.; Noguere, G.; Paris, M.; Pigni, M. T.; Plompen, A. J.; Pritychenko, B.; Pronyaev, V. G.; Roubtsov, D.; Rochman, D.; Romano, P.; Schillebeeckx, P.; Simakov, S.; Sin, M.; Sirakov, I.; Sleaford, B.; Sobes, V.; Soukhovitskii, E. S.; Stetcu, I.; Talou, P.; Thompson, I.; van der Marck, S.; Welser-Sherrill, L.; Wiarda, D.; White, M.; Wormald, J. L.; Wright, R. Q.; Zerkle, M.; Žerovnik, G.; Zhu, Y.

    2018-02-01

    We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.

  3. ENDF/B-VIII.0: The 8 th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D. A.; Chadwick, M. B.; Capote, R.; Kahler, A. C.; Trkov, A.; Herman, M. W.; Sonzogni, A. A.; Danon, Y.; Carlson, A. D.; Dunn, M.; Smith, D. L.; Hale, G. M.; Arbanas, G.; Arcilla, R.; Bates, C. R.; Beck, B.; Becker, B.; Brown, F.; Casperson, R. J.; Conlin, J.; Cullen, D. E.; Descalle, M. -A.; Firestone, R.; Gaines, T.; Guber, K. H.; Hawari, A. I.; Holmes, J.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Koning, A. J.; Kopecky, S.; Leal, L.; Lestone, J. P.; Lubitz, C.; Márquez Damián, J. I.; Mattoon, C. M.; McCutchan, E. A.; Mughabghab, S.; Navratil, P.; Neudecker, D.; Nobre, G. P. A.; Noguere, G.; Paris, M.; Pigni, M. T.; Plompen, A. J.; Pritychenko, B.; Pronyaev, V. G.; Roubtsov, D.; Rochman, D.; Romano, P.; Schillebeeckx, P.; Simakov, S.; Sin, M.; Sirakov, I.; Sleaford, B.; Sobes, V.; Soukhovitskii, E. S.; Stetcu, I.; Talou, P.; Thompson, I.; van der Marck, S.; Welser-Sherrill, L.; Wiarda, D.; White, M.; Wormald, J. L.; Wright, R. Q.; Zerkle, M.; Žerovnik, G.; Zhu, Y.

    2018-02-01

    We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.

  4. Measurement of thermal neutron cross-section and resonance integral for the 165Ho(n,γ) 166gHo reaction using electron linac-based neutron source

    Science.gov (United States)

    Nguyen, Van Do; Pham, Duc Khue; Kim, Tien Thanh; Kim, Guinyun; Lee, Manwoo; Kim, Kyung Sook; Kang, Heung-Sik; Cho, Moo-Hyun; Ko, In Soo; Namkung, Won

    2011-01-01

    The thermal neutron cross-section and the resonance integral of the 165Ho(n,γ) 166gHo reaction have been measured by the activation method using a 197Au(n,γ) 198Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation ( Fg), the thermal neutron self-shielding ( Gth), the resonance neutron self-shielding ( Gepi) effects, and the epithermal neutron spectrum shape factor ( α) were taken into account. The thermal neutron cross-section for the 165Ho(n,γ) 166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the 197Au(n,γ) 198Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the 165Ho(n,γ) 166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the 197Au(n,γ) 198Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.

  5. SPOTS, Library Generator for Program LEOPARD from Cross-Sections Data. LEOPARD, Fast and Thermal Neutron Spectra from Temperature and Geometry with Depletion Calculation

    International Nuclear Information System (INIS)

    Barry, R.F.; Krug, H.E P. Jr.

    1981-01-01

    1 - Description of problem or function: LEOPARD is a unit cell homogenization and spectrum generation (MUFT-SOFOCATE type) program with a fuel depletion option. 2 - Method of solution: The MUFT-SOFOCATE homogeneous medium spectrum analyses with heterogeneous corrections are used. The monoenergetic Amouyal-Benoist thermal disadvantage factor is applied at each of 172 SOFOCATE energy levels up to 0.625 eV. The U-238 resonance integral is forced to agree with a generalized Hellstrand correlation. 3 - Restrictions on the complexity of the problem: LEOPARD works with nuclides commonly used in water reactors. Thorium and U-238 fuel chains are allowed

  6. Actinide neutron-induced fission cross section measurements at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Tovesson, Fredrik K [Los Alamos National Laboratory; Laptev, Alexander B [Los Alamos National Laboratory; Hill, Tony S [INL

    2010-01-01

    Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications in a wide energy range from sub-thermal energies up to 200 MeV. A parallel-plate ionization chamber are used to measure fission cross sections ratios relative to the {sup 235}U standard while incident neutron energies are determined using the time-of-flight method. Recent measurements include the {sup 233,238}U, {sup 239-242}Pu and {sup 243}Am neutron-induced fission cross sections. Obtained data are presented in comparison with ex isting evaluations and previous data.

  7. Wind Turbine Radar Cross Section

    Directory of Open Access Journals (Sweden)

    David Jenn

    2012-01-01

    Full Text Available The radar cross section (RCS of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axis helical design, are shown. The unique electromagnetic scattering features, the effect of materials, and methods of mitigating wind turbine clutter are also discussed.

  8. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions; Modelo termohidraulico para realimentacao do calculo de secoes de choque neutronicas em reatores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Santiago, Daniela Maiolino Norberto

    2011-07-01

    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)

  9. Low energy total cross section of 36Ar

    International Nuclear Information System (INIS)

    Mughabghab, S.F.; Magurno, B.A.

    1975-01-01

    To compare the predictions of the valence model with measured partial radiative widths of 36 Ar an accurate knowledge of the bound-level parameters is required. This is achieved by carrying out a Breit-Wigner parameter fit to the total cross section of 36 Ar measured by Chrien et al and renormalized to the recommended values of the thermal capture and scattering cross sections. (1 figure, 1 table) (U.S.)

  10. Ideal gas scattering kernel for energy dependent cross-sections

    International Nuclear Information System (INIS)

    Rothenstein, W.; Dagan, R.

    1998-01-01

    A third, and final, paper on the calculation of the joint kernel for neutron scattering by an ideal gas in thermal agitation is presented, when the scattering cross-section is energy dependent. The kernel is a function of the neutron energy after scattering, and of the cosine of the scattering angle, as in the case of the ideal gas kernel for a constant bound atom scattering cross-section. The final expression is suitable for numerical calculations

  11. WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

    International Nuclear Information System (INIS)

    2004-01-01

    1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

  12. Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo [ed.

    1992-06-15

    Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.

  13. Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1992-06-01

    Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10 -5 eV to 20 MeV. Almost all the cross section data are reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in other tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum. (author)

  14. K+ nucleus total cross sections

    International Nuclear Information System (INIS)

    Sawafta, R.

    1990-01-01

    The scattering of K + mesons from nuclei has attracted considerable interest in the last few years. The K + holds a very special position as the weakest of all strongly interaction probes. The average cross section is not larger than about 10 mb at lab momenta below 800 MeV/c, corresponding to a mean free path in the nucleus larger than 5 fm. Thus the K + is capable of probing the entire volume of the nucleus. Single scattering of the K + with a nucleon in the nucleus dominates the nuclear scattering, and only small and calculable higher order corrections are needed. The nucleon is a dynamical entity and its internal structure can, in principle, be altered by its surrounding nuclear environment. This work reports an experiment in which the K + is used to compare the nucleon in the nucleus with a free nucleon

  15. Terahertz radar cross section measurements.

    Science.gov (United States)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-12-06

    We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar on full-size objects. The measurements are performed in a terahertz time-domain system with freely propagating terahertz pulses generated by tilted pulse front excitation of lithium niobate crystals and measured with sub-picosecond time resolution. The application of a time domain system provides ranging information and also allows for identification of scattering points such as weaponry attached to the aircraft. The shapes of the models and positions of reflecting parts are retrieved by the filtered back projection algorithm.

  16. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    1991-01-01

    In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months

  17. Implementation of the rapid cross section adjustment approach at General Electric

    International Nuclear Information System (INIS)

    Cowan, C.L.; Kujawski, E.; Protsik, R.

    1978-01-01

    The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

  18. Neutron cross sections: Book of curves

    International Nuclear Information System (INIS)

    McLane, V.; Dunford, C.L.; Rose, P.F.

    1988-01-01

    Neuton Cross Sections: Book of Curves represents the fourth edition of what was previously known as BNL-325, Neutron Cross Sections, Volume 2, CURVES. Data is presented only for (i.e., intergrated) reaction cross sections (and related fission parameters) as a function of incident-neutron energy for the energy range 0.01 eV to 200 MeV. For the first time, isometric state production cross sections have been included. 11 refs., 4 figs

  19. Nuclear Forensics and Radiochemistry: Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-08

    The neutron activation of components in a nuclear device can provide useful signatures of weapon design or sophistication. This lecture will cover some of the basics of neutron reaction cross sections. Nuclear reactor cross sections will also be presented to illustrate the complexity of convolving neutron energy spectra with nuclear excitation functions to calculate useful effective reactor cross sections. Deficiencies in the nuclear database will be discussed along with tools available at Los Alamos to provide new neutron cross section data.

  20. Porosity effects in the neutron total cross section of graphite

    International Nuclear Information System (INIS)

    Santisteban, J. R; Dawidowski, J; Petriw, S. N

    2009-01-01

    Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es

  1. Status of recent fast capture cross section evaluations for important fission product nuclides

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1982-01-01

    A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references

  2. [Absolute fission cross sections in the 14 MeV energy region]. Progress report, July 1982-June 1983

    International Nuclear Information System (INIS)

    1983-01-01

    Progress is reported on the following studies: thermal neutron absorption cross section of sulfur and the 252 Cf nu bar dilemma, the sigma (H)/sigma (Mn) cross section ratio, the sigma (H)/sigma (B) cross section ratio, 14 MeV neutron cross section measurements, beryllium-based pulsed neutron detector, and testing charged particle transport and Monte Carlo codes

  3. JENDL gas-production cross section file

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Narita, Tsutomu

    1992-05-01

    The JENDL gas-production cross section file was compiled by taking cross-section data from JENDL-3 and by using the ENDF-5 format. The data were given to 23 nuclei or elements in light nuclei and structural materials. Graphs of the cross sections and brief description on their evaluation methods are given in this report. (author)

  4. Integral nucleus-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Kumawat, H.

    2003-01-01

    Expressions approximating the experimental integral cross sections for elastic and inelastic interactions of light and heavy nuclei at the energies up to several GeV/nucleon are presented. The calculated cross sections are inside the corridor of experimental errors or very close to it. Described in detail FORTRAN code and a numerical example of the cross section approximation are also presented

  5. ERRFILS: a preliminary library of 30-group multigroup covariance data for use in CTR sensitivity studies

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Muir, D.W.

    1978-01-01

    A library of 30-group multigroup covariance data was prepared from preliminary ENDF/B-V data with the NJOY code. Data for Fe, Cr, Ni, 10 B, C, Cu, H, and Pb are included in this library. Reactions include total cross sections, elastic and inelastic scattering cross sections, and the most important absorption cross sections. Typical data from the file are shown. 3 tables

  6. Code implementation of partial-range angular scattering cross sections: GAMMER and MORSE

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1978-01-01

    A partial-range (finite-element) method has been previously developed for representing multigroup angular scattering in Monte Carlo photon transport. Computer application of the method, with preliminary quantitative results is discussed here. A multigroup photon cross section processing code, GAMMER, was written which utilized ENDF File 23 point data and the Klein--Nishina formula for Compton scattering. The cross section module of MORSE, along with several execution routines, were rewritten to permit use of the method with photon transport. Both conventional and partial-range techniques were applied for comparison to calculating angular and spectral penetration of 6-MeV photons through a six-inch iron slab. GAMMER was found to run 90% faster than SMUG, with further improvement evident for multiple-media situations; MORSE cross section storage was reduced by one-third; cross section processing, greatly simplified; and execution time, reduced by 15%. Particle penetration was clearly more forward peaked, as moment accuracy is retained to extremly high order. This method of cross section treatment offers potential savings in both storage and handling, as well as improved accuracy and running time in the actual execution phase. 3 figures, 4 tables

  7. BUGLE-93 (ENDF/B-VI) cross-section library data testing using shielding benchmarks

    International Nuclear Information System (INIS)

    Hunter, H.T.; Slater, C.O.; White, J.E.

    1994-01-01

    Several integral shielding benchmarks were selected to perform data testing for new multigroup cross-section libraries compiled from the ENDF/B-VI data for light water reactor (LWR) shielding and dosimetry. The new multigroup libraries, BUGLE-93 and VITAMIN-B6, were studied to establish their reliability and response to the benchmark measurements by use of radiation transport codes, ANISN and DORT. Also, direct comparisons of BUGLE-93 and VITAMIN-B6 to BUGLE-80 (ENDF/B-IV) and VITAMIN-E (ENDF/B-V) were performed. Some benchmarks involved the nuclides used in LWR shielding and dosimetry applications, and some were sensitive specific nuclear data, i.e. iron due to its dominant use in nuclear reactor systems and complex set of cross-section resonances. Five shielding benchmarks (four experimental and one calculational) are described and results are presented

  8. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    Knoll, G.F.

    1992-01-01

    From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase

  9. Plasma-based radar cross section reduction

    CERN Document Server

    Singh, Hema; Jha, Rakesh Mohan

    2016-01-01

    This book presents a comprehensive review of plasma-based stealth, covering the basics, methods, parametric analysis, and challenges towards the realization of the idea. The concealment of aircraft from radar sources, or stealth, is achieved through shaping, radar absorbing coatings, engineered materials, or plasma, etc. Plasma-based stealth is a radar cross section (RCS) reduction technique associated with the reflection and absorption of incident electromagnetic (EM) waves by the plasma layer surrounding the structure. A plasma cloud covering the aircraft may give rise to other signatures such as thermal, acoustic, infrared, or visual. Thus it is a matter of concern that the RCS reduction by plasma enhances its detectability due to other signatures. This needs a careful approach towards the plasma generation and its EM wave interaction. The book starts with the basics of EM wave interactions with plasma, briefly discuss the methods used to analyze the propagation characteristics of plasma, and its generatio...

  10. Relativistic photon-Maxwellian electron cross sections

    International Nuclear Information System (INIS)

    Wienke, B.R.; Lathrop, B.L.; Devaney, J.J.

    1986-01-01

    Temperature corrected cross sections, complementing the Klein-Nishina set, are developed for astrophysical, plasma, and transport applications. The set is obtained from a nonlinear least squares fit to the exact photon-Maxwellian electron cross sections, using the static formula as the asymptotic basis. Two parameters are sufficient (two decimal places) to fit the exact cross sections over a range of 0-100 keV in electron temperature, and 0-1 MeV in incident photon energy. The fit is made to the total cross sections, yet the parameters predict both total and differential scattering cross sections well. Corresponding differential energy cross sections are less accurate. An extended fit to (just) the total cross sections, over the temperature and energy range 0-5 MeV, is also described. (author)

  11. Fission cross-section normalization problems

    International Nuclear Information System (INIS)

    Wagemans, C.; Ghent Rijksuniversiteit; Deruytter, A.J.

    1983-01-01

    The present measurements yield σsub(f)-data in the neutron energy from 20 MeV to 30 keV directly normalized in the thermal region. In the keV-region these data are consistent with the absolute σsub(f)-measurements of Szabo and Marquette. For the secondary normalization integral I 2 values have been obtained in agreement with those of Gwin et al. and Czirr et al. which were also directly normalized in the thermal region. For the I 1 integral, however, puzzling low values have been obtained. This was also the case for σsub(f)-bar in neutron energy intervals containing strong resonances. Three additional measurements are planned to further investigate these observations: (i) maintaining the actual approx.2π-geometry but using a 10 B-foil for the neutron flux detection (ii) using a low detection geometry with a 10 B- as well as a 6 Li-flux monitor. Only after these measurements definite conclusions on the I 1 and I 2 integrals can be formulated and final σsub(f)-bar-values can be released. The present study also gives some evidence for a correlation between the integral I 2 and the neutron flux monitor used. The influence of a normalization via I 1 or I 2 on the final cross-section has been shown. The magnitude of possible normalization errors is illustrated. Finally, since 235 U is expected to be an ''easy'' nucleus (low α-activity high σsub(f)-values), there are some indications that the important discrepancies still present in 235 U(n,f) cross-section measurements might partially be due to errors in the neutron flux determination

  12. Cross-section libraries and kerma factors

    International Nuclear Information System (INIS)

    Little, R.C.; MacFarlane, R.E.; Seamon, R.E.

    1991-01-01

    A large amount of data is required in order to accurately simulate various aspects of Cold Neutron Sources using radiation transport codes such as MCNP and TWODANT. In particular, the following types of data are needed: couple neutron/photon transport libraries, neutron thermal S(α,β) data, response function data (including energy deposition), and proton interaction data. This paper concentrates on the coupled neutron/photon transport libraries and energy deposition. Data libraries available to radiation transport codes are obtained as a result of efforts in many areas, including differential and integral measurements, theoretical model codes, data evaluations, data processing, and data testing. A wide variety of data libraries are available to users of radiation transport codes, including pointwise and multigroup libraries. At Los Alamos, the authors generally recommend the use of data libraries derived from ENDF/B-V. It is often important to know how much energy is deposited in various regions of a device. This problem is typically modeled in radiation transport codes by folding the calculated fluences with an energy-dependent 'heating number'. The heating number represents the average energy deposited locally per collision. Calculation of these heating numbers from evaluated data libraries is fraught with difficulty. Many past difficulties related to energy deposition should be resolved by the release of ENDF/B-VI

  13. CHARTB multigroup transport package

    International Nuclear Information System (INIS)

    Baker, L.

    1979-03-01

    The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well

  14. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  15. Background-cross-section-dependent subgroup parameters

    International Nuclear Information System (INIS)

    Yamamoto, Toshihisa

    2003-01-01

    A new set of subgroup parameters was derived that can reproduce the self-shielded cross section against a wide range of background cross sections. The subgroup parameters are expressed with a rational equation which numerator and denominator are expressed as the expansion series of background cross section, so that the background cross section dependence is exactly taken into account in the parameters. The advantage of the new subgroup parameters is that they can reproduce the self-shielded effect not only by group basis but also by subgroup basis. Then an adaptive method is also proposed which uses fitting procedure to evaluate the background-cross-section-dependence of the parameters. One of the simple fitting formula was able to reproduce the self-shielded subgroup cross section by less than 1% error from the precise evaluation. (author)

  16. Scattering cross section for various potential systems

    Directory of Open Access Journals (Sweden)

    Myagmarjav Odsuren

    2017-08-01

    Full Text Available We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

  17. Scattering cross section for various potential systems

    Energy Technology Data Exchange (ETDEWEB)

    Odsuren, Myagmarjav; Khuukhenkhuu, Gonchigdorj; Davaa, Suren [Nuclear Research Center, School of Engineering and Applied Sciences, National University of Mongolia, Ulaanbaatar (Mongolia); Kato, Kiyoshi [Nuclear Reaction Data Centre, Faculty of Science, Hokkaido University, Sapporo (Japan)

    2017-08-15

    We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

  18. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1975-11-01

    The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

  19. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1976-05-01

    The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

  20. High ET jet cross sections at CDF

    International Nuclear Information System (INIS)

    Flaugher, B.

    1996-08-01

    The inclusive jet cross section for p anti p collisions at √s = 1.8 TeV as measured by the CDF collaboration will be presented. Preliminary CDF measurements of the Σ E T cross section at √s = 1.8 TeV and the central inclusive jet cross section at √s = 0.630 TeV will also be shown

  1. Measurements of neutron capture cross sections

    International Nuclear Information System (INIS)

    Nakajima, Yutaka

    1984-01-01

    A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238 U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)

  2. Recommended activation detector cross sections (RNDL-82)

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Lapenas, A.A.

    1984-01-01

    The results of the comparison between measured and calculated average cross sections in 5 benchmark experiments are presented. Calculations have been based on the data from 10 libraries of evaluated cross sections. The recommended library (RNDL-82) of the activation detector cross sections has been created on the basis of the comparison. RNDL-82, including 26 reactions, and the basic characteristics of the detectors are presented. (author)

  3. Impact of cross-section generation procedures on the simulation of the VVER 1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3-D thermal hydraulics/ neutron kinetics models

    International Nuclear Information System (INIS)

    Boyan D Ivanov; Kostadin N Ivanov; Sylvie Aniel; Eric Royer

    2005-01-01

    Full text of publication follows: In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D thermal hydraulics/neutron kinetics benchmark was defined. The overall objective OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3-D thermal hydraulics/ neutron kinetics models based on the data available in the benchmark specifications. The first code to code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between 2 sets of results, one of them being close to experimental results. The present paper focuses first on the analysis of the observed discrepancies. The VVER 1000 3-D thermal hydraulics/neutron kinetics models are based on thermal-hydraulic and neutronic data homogenized at the assembly scale. The neutronic data, provided as part of the benchmark specifications, consist thus in a set of parametrized 2 group cross sections libraries representing the different assemblies and the reflectors. The origin of the high observed discrepancies was found to lie in the use of these neutronic libraries. The concern was then to find a way to provide neutronic data, compatible with all the benchmark participants neutronic models, that enable also comparisons with experimental results. An analysis of the

  4. Curves and tables of neutron cross sections

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

    1990-07-01

    Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)

  5. Generation of neutron scattering cross sections for silicon dioxide

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  6. Electron collision cross sections of mercury

    International Nuclear Information System (INIS)

    Suzuki, Susumu; Kuzuma, Kiyotaka; Itoh, Haruo

    2006-01-01

    In this paper, we propose a new collision cross section set for mercury which revises the original set summarized by Hayashi in 1989. Hanne reported three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) determined from an electron beam experiment in 1988. As a matter for regret, no attentive consideration was given to combining these three excitation cross sections with the cross section set of Hayashi. Therefore we propose a new set where these three excitation cross sections are included. In this study, other two excitation cross sections (6 1 P 1 , 6 3 D 3 ) except for the three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) are taken from the original set of Hayashi. The momentum transfer cross section and the ionization collision cross section are also taken from Hayashi. A Monte Carlo Simulation (MCS) technique is applied for evaluating our new cross section set. The present results of the electron drift velocity and the ionization coefficient are compared to experimental values. Agreement is secured in relation to the electron drift velocity for 1.5 Td 2 ) is the reduced electric field, E (V/cm) is the electric field, N (1/cm 3 ) is the number density of mercury atoms at 0degC, 1 Torr, E/N is also equal to 2.828 x 10 -17 E/p 0 from the relation of the ideal gas equation, p 0 (Torr) is gas pressure at 0degC, 1 Torr=1.33322 x 10 -2 N/cm -2 and 10 -17 V/cm 2 is called 1 Td. Thus it is ensured that our new cross section set is reasonable enough to be used up to 100 eV when considering with the electron drift velocity and the ionization coefficient. (author)

  7. First measurement of the Rayleigh cross section

    NARCIS (Netherlands)

    Naus, H.; Ubachs, W.

    2000-01-01

    Rayleigh cross section for N2, Ar and SF6 was performed using the technique of cavity ring-down spectroscopy (CRDS). The experiment was based on the assumption that scattering cross section is equal to the extinction in the absence of absorption. The theory explains the molecular origin of

  8. Total cross section of highly excited strings

    International Nuclear Information System (INIS)

    Lizzi, F.; Senda, I.

    1990-01-01

    The unpolarized total cross section for the joining of two highly excited strings is calculated. The calculation is performed by taking the average overall states in the given excitation levels of the initial strings. We find that the total cross section grows with the energy and momentum of the initial states. (author). 8 refs, 1 fig

  9. Compilation of cross-sections. Pt. 1

    International Nuclear Information System (INIS)

    Flaminio, V.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

    1983-01-01

    A compilation of integral cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

  10. Total cross section results for deuterium electrodisintegration

    International Nuclear Information System (INIS)

    Skopik, D.M.; Murphy, J.J. II; Shin, Y.M.

    1976-01-01

    Theoretical total cross sections for deuterium electrodisintegration are presented as a function of incident electron energy. The cross section has been calculated using virtual photon theory with Partovi's photodisintegration calculation for E/subx/ > 10 MeV and effective range theory for E/subx/ 2 H(e, n) reaction in Tokamak reactors

  11. Model cross section calculations using LAHET

    International Nuclear Information System (INIS)

    Prael, R.E.

    1992-01-01

    The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented

  12. Vibrational enhancement of total breakup cross sections

    International Nuclear Information System (INIS)

    Haftel, M.I.; Lim, T.K.

    1984-01-01

    This paper considers the role of multi-two-body bound states, namely vibrational excitations, on total three-body breakup cross-sections. Total cross-sections are usually easy to measure, and they play a fundamental role in chemical kinetics. (orig.)

  13. Compilation of cross-sections. Pt. 4

    International Nuclear Information System (INIS)

    Alekhin, S.I.; Ezhela, V.V.; Lugovsky, S.B.; Tolstenkov, A.N.; Yushchenko, O.P.; Baldini, A.; Cobal, M.; Flaminio, V.; Capiluppi, P.; Giacomelli, G.; Mandrioli, G.; Rossi, A.M.; Serra, P.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

    1987-01-01

    This is the fourth volume in our series of data compilations on integrated cross-sections for weak, electromagnetic, and strong interaction processes. This volume covers data on reactions induced by photons, neutrinos, hyperons, and K L 0 . It contains all data published up to June 1986. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

  14. Penning ionization cross sections of excited rare gas atoms

    International Nuclear Information System (INIS)

    Ukai, Masatoshi; Hatano, Yoshihiko.

    1988-01-01

    Electronic energy transfer processes involving excited rare gas atoms play one of the most important roles in ionized gas phenomena. Penning ionization is one of the well known electronic energy transfer processes and has been studied extensively both experimentally and theoretically. The present paper reports the deexcitation (Penning ionization) cross sections of metastable state helium He(2 3 S) and radiative He(2 1 P) atoms in collision with atoms and molecules, which have recently been obtained by the authors' group by using a pulse radiolysis method. Investigation is made of the selected deexcitation cross sections of He(2 3 S) by atoms and molecules in the thermal collisional energy region. Results indicate that the cross sections are strongly dependent on the target molecule. The deexcitation probability of He(2 3 S) per collision increases with the excess electronic energy of He(2 3 S) above the ionization potential of the target atom or molecule. Another investigation, made on the deexcitation of He(2 1 P), suggests that the deexcitation cross section for He(2 1 P) by Ar is determined mainly by the Penning ionization cross section due to a dipole-dipole interaction. Penning ionization due to the dipole-dipole interaction is also important for deexcitation of He(2 1 P) by the target molecules examined. (N.K.)

  15. Recommended evaluation procedure for photonuclear cross section

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)

  16. Activation cross section data file, (1)

    International Nuclear Information System (INIS)

    Yamamuro, Nobuhiro; Iijima, Shungo.

    1989-09-01

    To evaluate the radioisotope productions due to the neutron irradiation in fission of fusion reactors, the data for the activation cross sections ought to be provided. It is planning to file more than 2000 activation cross sections at final. In the current year, the neutron cross sections for 14 elements from Ni to W have been calculated and evaluated in the energy range 10 -5 to 20 MeV. The calculations with a simplified-input nuclear cross section calculation system SINCROS were described, and another method of evaluation which is consistent with the JENDL-3 were also mentioned. The results of cross section calculation are in good agreement with experimental data and they were stored in the file 8, 9 and 10 of ENDF/B format. (author)

  17. Integral-capture measurements and cross-section adjustments for Nd, Sm, and Eu

    International Nuclear Information System (INIS)

    Anderl, R.A.; Schmittroth, F.; Harker, Y.D.

    1981-07-01

    Integral-capture reaction rates are reported for 143 Nd, 144 Nd, 145 Nd, 147 Sm, 151 Eu, 152 Eu, 153 Eu, and 154 Eu irradiated in different neutron spectra in EBR-II. These reaction rates are based primarily on mass-spectrometric measurements of the isotopic atom ratios of the capture product to the target nuclide. The neutron spectra are characterized using passive neutron dosimetry and spectrum-unfolding with the FERRET least-squares data analysis code. Reaction rates for the neutron spectrum monitors were determined by the radiometric technique using Ge(Li) spectrometers. These rates are also reported here. The integral data for the rare-earth samples and for the spectrum monitors were used in multigroup flux/cross-section adtustment analyses with FERRET to generate adjustments to 47 group representations of the ENDF/B-IV capture cross sections for the rare-earth isotopes. These adjusted cross sections are in good agreement with recent differential data and with adjusted cross sections based on STEK integral data. Examples are given of the use of the adjusted cross sections and covariance matrices for cross-section evaluation

  18. Is the quasielastic pion cross section really bigger than the pion-nucleus reaction cross section

    International Nuclear Information System (INIS)

    Silbar, R.R.

    1979-01-01

    It is shown that soft pion charge exchanges may increase the inclusive (π + ,π 0 ') cross section, relative to the total quasielastic (π + ,π + ') cross section, by as much as a factor of two. 4 references

  19. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

    The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

  20. Electron-impact cross sections of Ne

    International Nuclear Information System (INIS)

    Tsurubuchi, S.; Arakawa, K.; Kinokuni, S.; Motohashi, K.

    2000-01-01

    Electron-impact absolute emission cross sections were measured for the 3p→3s transitions of Ne. Excitation functions of the 3s→2p first resonance lines were measured in the energy range from the threshold to 1000 eV by a polarization-free optical method and relative cross sections were normalized to the absolute values, (41.0±5.4)x10 -19 cm 2 for the 73.6 nm line and (7.1±1.0)x10 -19 cm 2 for the 74.4 nm line, which were determined at 500 eV. The integrated level-excitation cross sections of Suzuki et al for the 1s 2 and 1s 4 levels were combined with the corresponding 3p→3s cascade cross sections obtained in this paper to give absolute emission cross sections for the resonance lines. The level-excitation cross sections of the 1s 2 and 1s 4 states in Paschen notation were determined from the threshold to 1000 eV by subtracting 3p→3s cascade cross sections from the corresponding 3s→2p emission cross sections of the resonance lines. A large cascade contribution is found in the emission cross section of the resonance lines. It is 28.5% for the 73.6 nm line and 49.6% for the 74.4 nm line at 40 eV, and 17.0 and 61.8%, respectively, at 300 eV. (author)

  1. Cross section for inelastic neutron acceleration by 178Hfm2

    International Nuclear Information System (INIS)

    Karamyan, S.A.; Carroll, J.J.

    2009-01-01

    The scattering of thermal neutrons from isomeric nuclei may include events in which the outgoing neutrons have increased kinetic energy. This process has been called Inelastic Neutron Acceleration (INNA) and occurs when the final nucleus after emission of the neutron is left in a state with lower energy than that of the isomer. The result, therefore, is an induced depletion of the isomeric population to the ground state. A cascade of several gammas must accompany the neutron emission to release the high angular momentum of the initial isomeric state. INNA was previously observed in a few cases and the associated cross sections were only in modest agreement with theoretical estimates. The most recent measurement of an INNA cross section was σ INNA = (258 ± 58) b for neutron scattering by 177 Lu m . In the present work, an INNA cross section of σ INNA = 152 -36 +51 b was deduced from measurements of the total burn-up of the high-spin, four-quasiparticle isomer 178 Hf m2 during irradiation by thermal neutrons. Statistical estimates for the probability of different reaction channels past neutron absorption were used in the analysis, and the deduced σ INNA is compared to the theoretically predicted cross section

  2. Cross-sectional anatomy for computed tomography

    International Nuclear Information System (INIS)

    Farkas, M.L.

    1988-01-01

    This self-study guide recognizes that evaluation and interpretation of CT-images demands a firm understanding of both cross-sectional anatomy and the principles of computed tomography. The objectives of this book are: to discuss the basic principles of CT, to stress the importance of cross-sectional anatomy to CT through study of selected cardinal transverse sections of head, neck, and trunk, to explain orientation and interpretation of CT-images with the aid of corresponding cross-sectional preparations

  3. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

    Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

    2010-06-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

  4. Differential Top Cross-section Measurements

    CERN Document Server

    Fenton, Michael James; The ATLAS collaboration

    2017-01-01

    The top quark is the heaviest known fundamental particle. The measurement of the differential top-quark pair production cross-section provides a stringent test of advanced perturbative QCD calculations. The ATLAS collaboration has performed detailed measurements of those differential cross sections at a centre-of-mass energy of 13 TeV. This talk focuses on differential cross-section measurements in the lepton+jets final state, including using boosted top quarks to probe our understanding of top quark production in the TeV regime.

  5. Hardon cross sections at ultra high energies

    International Nuclear Information System (INIS)

    Yodh, G.B.

    1987-01-01

    A review of results on total hadronic cross sections at ultra high energies obtained from a study of longitudinal development of cosmic ray air showers is given. The experimental observations show that proton-air inelastic cross section increases from 275 mb to over 500 mb as the collision energy in the center of mass increases from 20 GeV to 20 TeV. The proton-air inelastic cross section, obtained from cosmic ray data at √s = 30 TeV, is compared with calculations using various different models for the energy variation of the parameters of the elementary proton-proton interaction. Three conclusions are derived

  6. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

    Somogyi, G.; Bolzoni, P. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Trocsanyi, Z. [CERN, Geneva (Switzerland)

    2010-06-15

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

  7. NNLO jet cross sections by subtraction

    CERN Document Server

    Somogyi, Gabor; Trocsanyi, Zoltan

    2010-01-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of [1-4], over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  8. 12G: code for conversion of isotope-ordered cross-section libraries into group-ordered cross-section libraries

    International Nuclear Information System (INIS)

    Resnik, W.M. II; Bosler, G.E.

    1977-09-01

    Many current reactor physics codes accept cross-section libraries in an isotope-ordered form, convert them with internal preprocessing routines to a group-ordered form, and then perform calculations using these group-ordered data. Occasionally, because of storage and time limitations, the preprocessing routines in these codes cannot convert very large multigroup isotope-ordered libraries. For this reason, the I2G code, i.e., ISOTXS to GRUPXS, was written to convert externally isotope-ordered cross section libraries in the standard file format called ISOTXS to group-ordered libraries in the standard format called GRUPXS. This code uses standardized multilevel data management routines which establish a strategy for the efficient conversion of large libraries. The I2G code is exportable contingent on access to, and an intimate familiarization with, the multilevel routines. These routines are machine dependent, and therefore must be provided by the importing facility. 6 figures, 3 tables

  9. Calculation of the resonance cross section functions

    International Nuclear Information System (INIS)

    Slipicevic, K.F.

    1967-11-01

    This paper includes the procedure for calculating the Doppler broadened line shape functions ψ and χ which are needed for calculation of resonance cross section functions. The obtained values are given in tables

  10. Fission cross section measurements at intermediate energies

    International Nuclear Information System (INIS)

    Laptev, Alexander

    2005-01-01

    The activity in intermediate energy particle induced fission cross-section measurements of Pu, U isotopes, minor actinides and sub-actinides in PNPI of Russia is reviewed. The neutron-induced fission cross-section measurements are under way in the wide energy range of incident neutrons from 0.5 MeV to 200 MeV at the GNEIS facility. In number of experiments at the GNEIS facility, the neutron-induced fission cross sections were obtained for many nuclei. In another group of experiments the proton-induced fission cross-section have been measured for proton energies ranging from 200 to 1000 MeV at 100 MeV intervals using the proton beam of PNPI synchrocyclotron. (author)

  11. Measurement of multinucleon transfer cross-sections

    Indian Academy of Sciences (India)

    Keywords. Ni(C, ), Fe(C, ), =C, C, B, B, Be, Be, Be, Be, Li, Li; = 60 MeV; measured reaction cross-section; elastic scattering angular distribution; deduced transfer probabilities and enhancement factors.

  12. Capture cross sections on unstable nuclei

    Science.gov (United States)

    Tonchev, A. P.; Escher, J. E.; Scielzo, N.; Bedrossian, P.; Ilieva, R. S.; Humby, P.; Cooper, N.; Goddard, P. M.; Werner, V.; Tornow, W.; Rusev, G.; Kelley, J. H.; Pietralla, N.; Scheck, M.; Savran, D.; Löher, B.; Yates, S. W.; Crider, B. P.; Peters, E. E.; Tsoneva, N.; Goriely, S.

    2017-09-01

    Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

  13. Calculation of the resonance cross section functions

    Energy Technology Data Exchange (ETDEWEB)

    Slipicevic, K F [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-11-15

    This paper includes the procedure for calculating the Doppler broadened line shape functions {psi} and {chi} which are needed for calculation of resonance cross section functions. The obtained values are given in tables.

  14. Pion-nucleus cross sections approximation

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Polanski, A.; Sosnin, A.N.

    1990-01-01

    Analytical approximation of pion-nucleus elastic and inelastic interaction cross-section is suggested, with could be applied in the energy range exceeding several dozens of MeV for nuclei heavier than beryllium. 3 refs.; 4 tabs

  15. Status of neutron dosimetry cross sections

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.

    1992-01-01

    Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes

  16. Tachyonic ionization cross sections of hydrogenic systems

    Energy Technology Data Exchange (ETDEWEB)

    Tomaschitz, Roman [Department of Physics, Hiroshima University, 1-3-1 Kagami-yama, Higashi-Hiroshima 739-8526 (Japan)

    2005-03-11

    Transition rates for induced and spontaneous tachyon radiation in hydrogenic systems as well as the transversal and longitudinal ionization cross sections are derived. We investigate the interaction of the superluminal radiation field with matter in atomic bound-bound and bound-free transitions. Estimates are given for Ly-{alpha} transitions effected by superluminal quanta in hydrogen-like ions. The tachyonic photoelectric effect is scrutinized, in the Born approximation and at the ionization threshold. The angular maxima occur at different scattering angles in the transversal and longitudinal cross sections, which can be used to sift out longitudinal tachyonic quanta in a photon flux. We calculate the tachyonic ionization and recombination cross sections for Rydberg states and study their asymptotic scaling with respect to the principal quantum number. At the ionization threshold of highly excited states of order n {approx} 10{sup 4}, the longitudinal cross section starts to compete with photoionization, in recombination even at lower levels.

  17. a cross-sectional analytic study 2014

    African Journals Online (AJOL)

    Assessment of HIV/AIDS comprehensive correct knowledge among Sudanese university: a cross-sectional analytic study 2014. ... There are limited studies on this topic in Sudan. In this study we investigated the Comprehensive correct ...

  18. Capture cross sections on unstable nuclei

    Directory of Open Access Journals (Sweden)

    Tonchev A.P.

    2017-01-01

    Full Text Available Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

  19. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

    International Nuclear Information System (INIS)

    Dunn, M.E.; Greene, N.M.; Leal, L.C.

    1999-01-01

    Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

  20. Methods for calculating anisotropic transfer cross sections

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yixin.

    1985-01-01

    The Legendre moments of the group transfer cross section, which are widely used in the numerical solution of the transport calculation can be efficiently and accurately constructed from low-order (K = 1--2) successive partial range moments. This is convenient for the generation of group constants. In addition, a technique to obtain group-angle correlation transfer cross section without Legendre expansion is presented. (author)

  1. Heisenberg rise of total cross sections

    International Nuclear Information System (INIS)

    Ezhela, V.V.; Yushchenko, O.P.

    1988-01-01

    It is shown that on the basis of the original idea of Heisenberg on the quasiclassical picture of extended particle interactions one can construct a satisfactory description of the total cross sections, elastic cross sections, elastic diffractive slopes and mean charged multiplicities in the cm energy range from 5 to 900 GeV, and produce reasonable extrapolations up to several tens of TeV. 14 refs.; 7 figs.; 2 tabs

  2. Evaluation methods for neutron cross section standards

    International Nuclear Information System (INIS)

    Bhat, M.R.

    1980-01-01

    Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

    International Nuclear Information System (INIS)

    2005-01-01

    GMVP are based on the continuous-energy and multigroup method, respectively. In the continuous-energy method, all reactions are treated explicitly as given in evaluated nuclear data. Pointwise cross sections and angular/energy distributions are basically used for particle tracking. For neutron thermal scattering, the free gas model is used to take into account the thermal motion of target nuclei or the scattering law data S(α, β) and elastic thermal scattering representation in the ENDF format are used to take into account the binding effect in liquids and solids. In the unresolved resonance region of neutron cross sections, the probability table method is used. For photon reactions, detailed and simple models are available. The detailed model includes the generation of fluorescent X-rays in the photoelectric effect and the correction factor of the Klein-Nishina differential cross section for the incoherent scattering but the simple model does not include them. In both models, Bremsstrahlung photons can be optionally taken into account in the thick target approximation. Energy ranges are from 0.00005 eV to 20 MeV for neutrons and from 1 keV to 100 MeV for photons. In the multigroup method, all reactions are treated according to multigroup cross section data given by users

  5. Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-12-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt

  6. Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1991-01-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)

  7. Performance of JEF2.2 based continuous energy cross sections in predicting the multiplication factor of critical systems

    International Nuclear Information System (INIS)

    John, T.M.; de Leege, P.F.A.; Hoogenboom, J.E.

    1996-01-01

    The continuous energy representation of cross sections for neutronics calculations avoids the requirement of resonance self shielding and the assumptions about the neutron spectrum used for weighing cross sections, required in the preparation of a multigroup cross sections library. The cross sections library prepared for a particular temperature of the nuclide is valid irrespective of the environment of the nuclide and can be used in calculations for many types of reactors. It is comparatively easier to incorporate them in Monte Carlo simulation of neutron transport. The Monte Carlo code MCNP is capable of using a continuous energy representation of nuclear cross sections in simulation of neutron or photon transport. The ACER module of NJOY is able to generate the continuous energy cross section of any nuclide in a format that can be used by MCNP, from any evaluated data file in ENDF/B format. Continuous energy cross sections prepared from the evaluated data file JEF2.2 was used to analyse some standard critical benchmarks and also the critical configuration of the HOR, a 2 MW research reactor at Delft, the Netherlands. Results show that continuous energy cross sections prepared from JEF2.2 evaluated file predicts the multiplication factor of critical systems very close to unity. (author). 6 refs., 2 tabs., 1 fig

  8. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

    Hudson, Nathanael H.; Ougouag, Abderrafi M.; Rahnema, Farzad; Gougar, Hans

    2009-01-01

    A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.

  9. Effective cross sections of U-235 and Au in a TRIGA-type reactor core

    International Nuclear Information System (INIS)

    Harasawa, S.; Auu, G.A.

    1992-01-01

    The dependence of effective cross sections of gold and uranium for neutron spectrum in Rikkyo University Reactor (TRIGA Mark- II, RUR) fuel cell was studied using computer calculations. The dependence of thermal neutron spectrum with temperature was also investigated. The effective cross section of gold in water of the fuel cell at 32degC was 90.3 barn and the fission cross section of U-235, 483 barn. These two values are similar to the cross sections for neutron energy of 0.034 eV. (author)

  10. Creation and validation of a neutron-gamma coupled multigroup cross section library

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Lee, S.M.

    1995-01-01

    The task of creating our own neutron-gamma coupled library was taken up. By using 1985 version of NJOY code system, a coupled set called IGC-DE4-S1 in ANISN format for 25 nuclides has been arrived at based on ENDF/B-IV neutron library and DLC-99 gamma library, with Legendre order of up to 5. The flow chart for the creation of coupled set is given. 5 refs, 1 fig., 3 tabs

  11. Generation of multigroup cross sections from ENDF/B-IV nuclear data library

    International Nuclear Information System (INIS)

    Chapot, J.L.C.; Thome Filho, Z.D.

    1980-04-01

    The generation of nuclear data compacted in energy groups is made. The nuclear data library ENDF/B-IV, Evaluated Nuclear Data File, and the new version of the codes ETOG-3 and ETOT-3 are utilized. The data obtained are compared with data from other sources. (L.F.) [pt

  12. Neutron capture cross section standards for BNL-325

    International Nuclear Information System (INIS)

    Holden, N.E.

    1980-01-01

    The most common cross section standards for capture reactions in the thermal neutron energy region are gold, cobalt, and manganese. In preparation for the fourth edition of BNL-325, data on the thermal cross section and resonance integral were evaluated for these three standards. For gold, only measurements below the Bragg scattering cutoff were used and extrapolated to a neutron velocity of 2200 meters/second. A non 1/v correction due to the 4.9 eV resonance was made. The resonance integral is based on Jirlow's integral measurement and Tellier's parameters. The resonance integrals for cobalt and manganese are based solely on integral measurements because the capture widths of the first major resonance either vary by 20% in various measurements (cobalt), or have never been measured (manganese). Recommended thermal cross sections and resonance integrals are respectively gold: 98.65/plus or minus/0.9 barns, 1550/plus or minus/28 barns; cobalt: 37.18/plus or minus/0.06 barns, 74.2/plus or minus/2.0 barns and manganese: 13.3/plus or minus/0.2 barns, and 14.0/plus or minus/0.3 barns. 72 refs

  13. 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)

  14. Top quark production cross-section measurements

    CERN Document Server

    Chen, Ye; The ATLAS collaboration

    2017-01-01

    Measurements of the inclusive and differential cross-sections for top-quark pair and single top production cross sections in proton-proton collisions with the ATLAS detector at the Large Hadron Collider are presented at center-of-mass energies of 8 TeV and 13 TeV. The inclusive measurements reach high precision and are compared to the best available theoretical calculations. These measurements, including results using boosted tops, probe our understanding of top-pair production in the TeV regime. The results are compared to Monte Carlo generators implementing LO and NLO matrix elements matched with parton showers and NLO QCD calculations. For the t-channel single top measurement, the single top-quark and anti-top-quark total production cross-sections, their ratio, as well as differential cross sections are also presented. A measurement of the production cross-section of a single top quark in association with a W boson, the second largest single-top production mode, is also presented. Finally, measurements of ...

  15. Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  16. Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  17. Homogenized group cross sections by Monte Carlo

    International Nuclear Information System (INIS)

    Van Der Marck, S. C.; Kuijper, J. C.; Oppe, J.

    2006-01-01

    Homogenized group cross sections play a large role in making reactor calculations efficient. Because of this significance, many codes exist that can calculate these cross sections based on certain assumptions. However, the application to the High Flux Reactor (HFR) in Petten, the Netherlands, the limitations of such codes imply that the core calculations would become less accurate when using homogenized group cross sections (HGCS). Therefore we developed a method to calculate HGCS based on a Monte Carlo program, for which we chose MCNP. The implementation involves an addition to MCNP, and a set of small executables to perform suitable averaging after the MCNP run(s) have completed. Here we briefly describe the details of the method, and we report on two tests we performed to show the accuracy of the method and its implementation. By now, this method is routinely used in preparation of the cycle to cycle core calculations for HFR. (authors)

  18. Partial cross sections near the higher resonances

    International Nuclear Information System (INIS)

    Falk-Vairant, P.; Valladas, G.

    1961-07-01

    As a continuation of the report given at the 10. Rochester Conference, recent measurements of charge-exchange cross section and π 0 production in π - -p interactions are presented here. Section 1 gives a summary of the known results for the elastic, inelastic, and charge-exchange cross sections. Section 2 presents the behavior of the cross sections in the T=1/2 state, in order to discuss the resonances at 600 and 890 MeV. Section 3 discusses the charge-exchange scattering and the interference term between the T=1/2 and T=3/2 states. Section 4 presents some comments on inelastic processes. This report is reprinted from 'Reviews of Modern Physics', Vol. 33, No. 3, 362-367, July, 1961

  19. Photoproton cross section for /sup 19/F

    Energy Technology Data Exchange (ETDEWEB)

    Tsubota, H [Tohoku Univ., Sendai (Japan). Coll. of General Education; Kawamura, N; Oikawa, S; Uegaki, J I

    1975-02-01

    Proton energy spectra have been measured at 90/sup 0/ for the /sup 19/F(e,e'p)/sup 18/O reaction in the giant resonance region. The (..gamma..,p/sub 0/) and (..gamma..,p/sub 1/) differential cross sections are extracted from the proton energy spectra by using virtual-photon spectra. The integrated differential cross section of the (..gamma..,p/sub 0/) and (..gamma..,p/sub 1/) reactions are 1.80+-0.27 and 0.50+-0.45 MeV-mb/sr, respectively. The results are discussed with the shell model theory by comparing with the (..gamma..,p/sub 0/) cross section of the neighboring 4n-nucleus /sup 20/Ne. A significant increase of the proton yield leaving the non-ground states is found at 25 MeV of the incident electron energy. This is discussed in terms of the core excitation effect.

  20. Prospects for Precision Neutrino Cross Section Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Deborah A. [Fermilab

    2016-01-28

    The need for precision cross section measurements is more urgent now than ever before, given the central role neutrino oscillation measurements play in the field of particle physics. The definition of precision is something worth considering, however. In order to build the best model for an oscillation experiment, cross section measurements should span a broad range of energies, neutrino interaction channels, and target nuclei. Precision might better be defined not in the final uncertainty associated with any one measurement but rather with the breadth of measurements that are available to constrain models. Current experience shows that models are better constrained by 10 measurements across different processes and energies with 10% uncertainties than by one measurement of one process on one nucleus with a 1% uncertainty. This article describes the current status of and future prospects for the field of precision cross section measurements considering the metric of how many processes, energies, and nuclei have been studied.

  1. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

    Somogyi, G.; Bolzoni, P. [DESY, Platanenallee 6, D-15738 Zeuthen (Germany); Trocsanyi, Z. [CERN PH-TH, on leave from University of Debrecen and Institute of Nuclear Research of HAS, H-4001 P.O.Box 51 (Hungary)

    2010-08-15

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  2. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

    Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

    2010-01-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  3. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

  4. Optical Model and Cross Section Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.W.; Pigni, M.T.; Dietrich, F.S.; Oblozinsky, P.

    2009-10-05

    Distinct minima and maxima in the neutron total cross section uncertainties were observed in model calculations using spherical optical potential. We found this oscillating structure to be a general feature of quantum mechanical wave scattering. Specifically, we analyzed neutron interaction with 56Fe from 1 keV up to 65 MeV, and investigated physical origin of the minima.We discuss their potential importance for practical applications as well as the implications for the uncertainties in total and absorption cross sections.

  5. Cross sections for charm production by neutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Ushida, N [Aichi Univ. of Education, Kariya (Japan). Dept. of Physics; Kondo, T [Fermi National Accelerator Lab., Batavia, IL (USA); Fujioka, G; Fukushima, J; Takahashi, Y; Tatsumi, S; Yokoyama, C [Kobe Univ. (Japan). Dept. of Physics; Homma, Y; Tsuzuki, Y [Kobe Univ. (Japan). Coll. of Liberal Arts; Bahk, S

    1983-02-03

    The production of charmed particles has been measured using a hybrid emulsion spectrometer in the Fermilab wide-band neutrino beam. The relative cross section for charged current charmed particle production is sigma(v -> ..mu../sup -/c)/sigma(v -> ..mu../sup -/) = 6.5 +- 1.9/1.8%, and the energy dependence of the cross section is presented. One event with charm pair production was observed. A limit of sigma(v -> ..mu..canti c)/sigma(v -> ..mu..c) < 6% (90% CL) is found for the ratio of charged current pair and single charm production.

  6. Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

    International Nuclear Information System (INIS)

    Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

  7. 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

  8. Neutron total cross section measurements on 249Cf

    International Nuclear Information System (INIS)

    Carlton, R.F.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Benjamin, R.W.

    1979-01-01

    Neutron total cross section measurements were performed on a sample of 249 Cf (5.65 mg total weight) with the ORELA as a source of pulsed neutrons. The sample, the inverse thickness of which was 1542 barns/atom, consisted of 85.3% 249 Cf and 14.4% 249 Bk, and was cooled to liquid nitrogen temperature. Analyses were also made of data from a thin sample (l/n = 17430) of 65% 249 Cf in the region of the large fission resonance at 0.7 eV. Fifty-five resonances in 249 Cf were observed and analyzed over the energy range 0.1 eV to 90 eV by use of an R-matrix multilevel formalism. The resonance parameters obtained were used to determine the level spacing and the s-wave neutron and fission strength functions. Thermal total cross section measurements were also performed. 5 figures, 3 tables

  9. Electron attachment cross sections obtained from electron attachment spectroscopy

    International Nuclear Information System (INIS)

    Popp, P.; Baumbach, J.I.; Leonhardt, J.W.; Mothes, S.

    1988-01-01

    Electron capture detectors have a high sensitivity for substances with high thermal electron attachment cross sections. The electron attachment spectroscopy makes it possible to change the mean electron energy in such a way that the maximum for dissociative electron attachment is reached. Thus, best operation modes of the detection system as well as significant dependencies of electron attachment coefficients are available. Cross sections for electron attachment as a function of the electron energy are obtained with the knowledge of electron energy distribution functions from Boltzmann equation analysis by a special computer code. A disadvantage of this electron attachment spectroscopy is the superposition of space charge effects due to the decrease of the electron drift velocity with increasing mean electron energy. These influences are discussed. (author)

  10. AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

    It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

  11. Ecological Panel Inference from Repeated Cross Sections

    NARCIS (Netherlands)

    Pelzer, Ben; Eisinga, Rob; Franses, Philip Hans

    2004-01-01

    This chapter presents a Markov chain model for the estimation of individual-level binary transitions from a time series of independent repeated cross-sectional (RCS) samples. Although RCS samples lack direct information on individual turnover, it is demonstrated here that it is possible with these

  12. Validation of evaluated neutron standard cross sections

    International Nuclear Information System (INIS)

    Badikov, S.; Golashvili, T.

    2008-01-01

    Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)

  13. Stability of tokamaks with elongated cross section

    International Nuclear Information System (INIS)

    An, C.H.; Bateman, G.

    1978-08-01

    Fixed boundary n = 1 MHD instabilities are studied computationally as a function of diamagnetism (β/sub pol/) and current profile in elongated toroidal equilibria (1 2) or a diamagnetic plasma (β/sub pol/ > 1) with only a mildly elongated cross section

  14. Photoelectric absorption cross sections with variable abundances

    Science.gov (United States)

    Balucinska-Church, Monika; Mccammon, Dan

    1992-01-01

    Polynomial fit coefficients have been obtained for the energy dependences of the photoelectric absorption cross sections of 17 astrophysically important elements. These results allow the calculation of X-ray absorption in the energy range 0.03-10 keV in material with noncosmic abundances.

  15. (, 3) Differential cross section of He

    Indian Academy of Sciences (India)

    The angular distribution of the five-fold differential cross section for the electron impact double ionization of He (21 ) and He (23 ) has been studied. The kinematical conditions for maxima/minima in the angular distribution for the two cases have been compared. The two-step process for the double ionization is found to ...

  16. Precise relative cross sections for np scattering

    International Nuclear Information System (INIS)

    Goetz, J.; Brogli-Gysin, C.; Hammans, M.; Haffter, P.; Henneck, R.; Jourdan, J.; Masson, G.; Qin, L.M.; Robinson, S.; Sick, I.; Tuccillo, M.

    1994-01-01

    We present data on the differential cross section for neutron-proton scattering for an incident neutron energy of 67 MeV. These data allow a precise determination of the 1 P 1 phase which, in phase-shift analyses, is strongly correlated with the S-D amplitude which we are measuring via different observables. ((orig.))

  17. Symmetric charge transfer cross section of uranium

    International Nuclear Information System (INIS)

    Shibata, Takemasa; Ogura, Koichi

    1995-03-01

    Symmetric charge transfer cross section of uranium was calculated under consideration of reaction paths. In the charge transfer reaction a d 3/2 electron in the U atom transfers into the d-electron site of U + ( 4 I 9/2 ) ion. The J value of the U atom produced after the reaction is 6, 5, 4 or 3, at impact energy below several tens eV, only resonant charge transfer in which the product atom is ground state (J=6) takes place. Therefore, the cross section is very small (4-5 x 10 -15 cm 2 ) compared with that considered so far. In the energy range of 100-1000eV the cross section increases with the impact energy because near resonant charge transfer in which an s-electron in the U atom transfers into the d-electron site of U + ion. Charge transfer cross section between U + in the first excited state (289 cm -1 ) and U in the ground state was also obtained. (author)

  18. LAMBDA p total cross-section measurement

    CERN Multimedia

    CERN PhotoLab

    1970-01-01

    A view of the apparatus used for the LAMBDA p total cross-section measurement at the time of its installation. The hyperons decaying into a proton and a pion in the conical tank in front were detected in the magnet spectrometer in the upper half of the picture. A novel detection technique using exclusively multiwire proportional chambers was employed.

  19. Measurement cross sections for radioisotopes production

    International Nuclear Information System (INIS)

    Garrido, E.

    2011-01-01

    New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β - - 47 Sc, 67 Cu - β + - 44 Sc, 64 Cu, 82 Sr/ 82 Rb, 68 Ge/ 68 Ga - and α emitters - 211 At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - nat Cu or nat Ni - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the nat Ti(p,X) 47 Sc and 68 Zn(p,2p) 67 Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68 Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)

  20. Experimental determination of resonance absorption cross sections for Zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1968-05-15

    The integral absorption cross section for the neutron spectrum and the thermal absorption cross section for zircaloy-2 have been determined using the pile oscillator technique. Using both values and a measured ratio of the epithermal to the thermal flux, the effective resonance integrals were obtained. After subtraction of the contributions for alloy and impurity elements, the effective resonance integrals for zirconium were evaluated. An extrapolated value of 0.91{+-}0.10 was obtained for the dilute integral. (author)

  1. Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

  2. Stellar neutron capture cross sections of the Ba isotopes

    International Nuclear Information System (INIS)

    Voss, F.; Wisshak, K.; Guber, K.; Kaeppeler, F.; Reffo, G.

    1994-03-01

    The neutron capture cross sections of 134 Ba, 135 Ba, 136 Ba, and 137 Ba were measured in the energy range from 5 to 225 keV at the Karlsruhe 3.75 MV Van de Graaff accelerator. Neutrons were produced via the 7 Li(p,n) 7 Be reaction by bombarding metallic Li targets with a pulsed proton beam. Capture events were registered with the Karlsruhe 4π Barium Fluoride Detector. Several runs have been performed under different experimental conditions to study the systematic uncertainties, which resulted mainly from the large ratios of total to capture cross sections of up to 400. The cross section ratios were determined with an overall uncertainty of ∼3%, an improvement by factors of five to eight compared to existing data. Severe discrepancies were found with respect to previous results. Maxwellian averaged neutron capture cross sections were calculated for thermal energies between kT=10 keV and 100 keV. These stellar cross sections were used in an s-process analysis. For the s-only isotopes 134 Ba and 136 Ba the N s ratio was determined to 0.875±0.025. Hence, a significant branching of the s-process path at 134 Cs can be claimed for the first time, in contrast to predictions from the classical approach. This branching yields information on the s-process temperature, indicating values around T 8 =2. The new cross sections are also important for the interpretation of barium isotopic anomalies, which were recently discovered in SiC grains of carbonaceous chondrite meteorites. Together with the results from previous experiments on tellurium and samarium, a general improvement of the N s systematics in the mass range A=120 to 150 is achieved. This allows for a more reliable separation of s- and r-process yields, resulting in an improved assignment of the respective contributions to elemental barium that is required for comparison with stellar observations. (orig.) [de

  3. Evaluation of cross-section uncertainties using physical constraints for 238U, 239Pu

    International Nuclear Information System (INIS)

    De Saint Jean, Cyrille; Privas, Edwin; Archier, Pascal; Noguere, Gilles; Litaize, Olivier; Leconte, Pierre; Bernard, David

    2014-01-01

    Neutron-induced reactions between 0 eV and 20 MeV are based on various physical properties such as nuclear reaction models, microscopic and integral measurements. Most of the time, the evaluation work is done independently between the resolved resonance range and the continuum, giving rise to mismatches for the cross-sections, larger uncertainties on boundary and no cross-correlation between high-energy domain and resonance range. In addition the use of integral experiment is sometimes only related to central values (evaluation is 'working fine' on a dedicated set of benchmarks) and reductions of uncertainties are not straightforward on cross-sections themselves: working fine could be mathematically reflected by a reduced uncertainty. As the CIELO initiative is to bring experts in each field to propose/discuss these matters, after having presented the status of 238 U and 239 Pu cross-sections covariances evaluation (for JEFF-3.2 as well as the WPEC SG34 subgroup), this paper will present several methodologies that may be used to avoid such effects on covariances. A first idea based on the use of experiments overlapping two energy domains appeared in the near past. It was reviewed and extended to the use of systematic uncertainties (normalisation for example) and for integral experiments as well. In addition, we propose a methodology taking into account physical constraints on an overlapping energy domain where both nuclear reaction models are used (continuity of both cross-sections and derivatives for example). The use of Lagrange multiplier (related to these constraints) in a classical generalised least square procedure will be exposed. Some academic examples will then be presented for both point-wise and multi-group cross-sections to present the methodologies. In addition, new results for 239 Pu will be presented on resonance range and higher energies to reduce capture and fission cross-section uncertainties by using integral experiments (JEZEBEL experiment as

  4. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.

  5. (n,{alpha}) cross section measurement of gaseous sample using gridded ionization chamber. Cross section determination

    Energy Technology Data Exchange (ETDEWEB)

    Sanami, Toshiya; Baba, Mamoru; Saito, Keiichiro; Ibara, Yasutaka; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1997-03-01

    We are developing a method of (n,{alpha}) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the {sup 12}C(n,{alpha}{sub 0}){sup 9}Be measurement. We applied this method to the {sup 16}O(n,{alpha}){sup 13}C cross section around 14.1 MeV. (author)

  6. Summary of the Workshop on Neutron Cross Section Covariances

    International Nuclear Information System (INIS)

    Smith, Donald L.

    2008-01-01

    A Workshop on Neutron Cross Section Covariances was held from June 24-27, 2008, in Port Jefferson, New York. This Workshop was organized by the National Nuclear Data Center, Brookhaven National Laboratory, to provide a forum for reporting on the status of the growing field of neutron cross section covariances for applications and for discussing future directions of the work in this field. The Workshop focused on the following four major topical areas: covariance methodology, recent covariance evaluations, covariance applications, and user perspectives. Attention was given to the entire spectrum of neutron cross section covariance concerns ranging from light nuclei to the actinides, and from the thermal energy region to 20 MeV. The papers presented at this conference explored topics ranging from fundamental nuclear physics concerns to very specific applications in advanced reactor design and nuclear criticality safety. This paper provides a summary of this workshop. Brief comments on the highlights of each Workshop contribution are provided. In addition, a perspective on the achievements and shortcomings of the Workshop as well as on the future direction of research in this field is offered

  7. Rotational averaging of multiphoton absorption cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Friese, Daniel H., E-mail: daniel.h.friese@uit.no; Beerepoot, Maarten T. P.; Ruud, Kenneth [Centre for Theoretical and Computational Chemistry, University of Tromsø — The Arctic University of Norway, N-9037 Tromsø (Norway)

    2014-11-28

    Rotational averaging of tensors is a crucial step in the calculation of molecular properties in isotropic media. We present a scheme for the rotational averaging of multiphoton absorption cross sections. We extend existing literature on rotational averaging to even-rank tensors of arbitrary order and derive equations that require only the number of photons as input. In particular, we derive the first explicit expressions for the rotational average of five-, six-, and seven-photon absorption cross sections. This work is one of the required steps in making the calculation of these higher-order absorption properties possible. The results can be applied to any even-rank tensor provided linearly polarized light is used.

  8. Measurements of neutron spallation cross section. 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, E.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Imamura, M.; Nakao, N.; Shibata, S.; Uwamino, Y.; Nakanishi, N.; Tanaka, Su.

    1997-03-01

    Neutron spallation cross section of {sup 59}Co(n,xn){sup 60-x}Co, {sup nat}Cu(n,sp){sup 56}Mn, {sup nat}Cu(n,sp){sup 58}Co, {sup nat}Cu(n,xn){sup 60}Cu, {sup nat}Cu(n,xn){sup 61}Cu and {sup nat}Cu(n,sp){sup 65}Ni was measured in the quasi-monoenergetic p-Li neutron fields in the energy range above 40 MeV which have been established at three AVF cyclotron facilities of (1) INS of Univ. of Tokyo, (2) TIARA of JAERI and (3) RIKEN. Our experimental data were compared with the ENDF/B-VI high energy file data by Fukahori and the calculated cross section data by Odano. (author)

  9. Structured ion impact: Doubly differential cross sections

    International Nuclear Information System (INIS)

    DuBois, R.D.

    1987-01-01

    The electron emission in coincidence with a projectile that has been ionized has been measured, thus making it possible to separate and identify electrons resulting from these various mechanisms. In 1985, coincidence doubly differential cross sections were measured for 400 to 750 keV/atomic mass unit (amu) He + impact on He, Ne, Ar, Kr, and H 2 O. Cross sections were measured for selected angles and for electron energies ranging from 10 to 1000 eV. Because of the coincidence mode of measurement, the total electron emission was subdivided into its target emission and its projectile emission components. The most interesting findings were that target ionization does not account for the electron emission spectrum at lower electron energies. A sizable percentage of these low-energy electrons were shown to originate as a result of simultaneous projectile/target ionizations. Similar features were observed for all targets and impact energies that were studied

  10. Electron-collision cross sections for iodine

    International Nuclear Information System (INIS)

    Zatsarinny, O.; Bartschat, K.; Garcia, G.; Blanco, F.; Hargreaves, L.R.; Jones, D.B.; Murrie, R.; Brunton, J.R.; Brunger, M.J.; Hoshino, M.; Buckman, S.J.

    2011-01-01

    We present results from a joint experimental and theoretical study of elastic electron scattering from atomic iodine. The experimental results were obtained by subtracting known cross sections from the measured data obtained with a pyrolyzed mixed beam containing a variety of atomic and molecular species. The calculations were performed using both a fully relativistic Dirac B-spline R-matrix (close-coupling) method and an optical model potential approach. Given the difficulty of the problem, the agreement between the two sets of theoretical predictions and the experimental data for the angle-differential and the angle-integrated elastic cross sections at 40 eV and 50 eV is satisfactory.

  11. Absolute partial photoionization cross sections of ethylene

    Science.gov (United States)

    Grimm, F. A.; Whitley, T. A.; Keller, P. R.; Taylor, J. W.

    1991-07-01

    Absolute partial photoionization cross sections for ionization out of the first four valence orbitals to the X 2B 3u, A 2B 3g, B 2A g and C 2B 2u states of the C 2H 4+ ion are presented as a function of photon energy over the energy range from 12 to 26 eV. The experimental results have been compared to previously published relative partial cross sections for the first two bands at 18, 21 and 24 eV. Comparison of the experimental data with continuum multiple scattering Xα calculations provides evidence for extensive autoionization to the X 2B 3u state and confirms the predicted shape resonances in ionization to the A 2B 3g and B 2A g states. Identification of possible transitions for the autoionizing resonances have been made using multiple scattering transition state calculations on Rydberg excited states.

  12. Radar cross section measurements using terahertz waves

    DEFF Research Database (Denmark)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-01-01

    Radar cross sections at terahertz frequencies are measured on scale models of aircrafts. A time domain broadband THz system generates freely propagating THz pulses measured with sub-picosecond time resolution. The THz radiation is generated using fs laser pulses by optical rectification...... in order to measure realistic radar cross sections. RCS polar and azimuthal angle plots of F-16 and F-35 are presented....... in a lithium niobate crystal with application of the tilted wave front method, resulting in high electric field THz pulses with a broad band spectrum from 100 GHz up to 4 THz. The corresponding wave lengths are two orders of magnitude smaller than normal radars and we therefore use scale models of size 5-10 cm...

  13. Electron Capture Cross Sections for Stellar Nucleosynthesis

    Directory of Open Access Journals (Sweden)

    P. G. Giannaka

    2015-01-01

    Full Text Available In the first stage of this work, we perform detailed calculations for the cross sections of the electron capture on nuclei under laboratory conditions. Towards this aim we exploit the advantages of a refined version of the proton-neutron quasiparticle random-phase approximation (pn-QRPA and carry out state-by-state evaluations of the rates of exclusive processes that lead to any of the accessible transitions within the chosen model space. In the second stage of our present study, we translate the abovementioned e--capture cross sections to the stellar environment ones by inserting the temperature dependence through a Maxwell-Boltzmann distribution describing the stellar electron gas. As a concrete nuclear target we use the 66Zn isotope, which belongs to the iron group nuclei and plays prominent role in stellar nucleosynthesis at core collapse supernovae environment.

  14. Test of RIPL-2 cross section calculations

    International Nuclear Information System (INIS)

    Herman, M.

    2002-01-01

    The new levels and optical segments and microscopic HF-BCS level densities (part of the density segment) were tested in practical calculations of cross sections for neutron induced reactions on 22 targets (40-Ca, 47-Ti, 52-Cr, 55-Mn, 58-Ni, 63-Cu, 71-Ga, 80-Se, 92-Mo, 93-Nb, 100-Mo, 109-Ag, 114-Cd, 124-Sn, 127-I, 133-Cs, 140-Ce, 153-Eu, 169-Tm, 186-W, 197-Au, 208-Pb). For each target all reactions involving up to 3 neutron, 1 proton and 1 α-particle emissions (subject to actual reaction thresholds) were considered in the incident energy range from 1 keV up to 20 MeV (in some cases up to 27 MeV). In addition, total, elastic, and neutron capture cross sections were calculated

  15. Double differential cross sections of ethane molecule

    Science.gov (United States)

    Kumar, Rajeev

    2018-05-01

    Partial and total double differential cross sections corresponding to various cations C2H6+, C2H4+, C2H5+, C2H3+, C2H2+, CH3+, H+, CH2+, C2H+, H2+, CH+, H3+, C2+ and C+ produced during the direct and dissociative electron ionization of Ethane (C2H6) molecule have been calculated at fixed impinging electron energies 200 and 500eV by using modified Jain-Khare semi empirical approach. The calculation for double differential cross sections is made as a function of energy loss suffered by primary electron and angle of incident. To the best of my knowledge no other data is available for the comparison.

  16. Cross sections required for FMIT dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.; Lippincott, E.P.; Mann, F.M.; Oberg, D.L.; Roberts, J.H.; Ruddy, F.H.

    1980-01-01

    The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies

  17. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  18. Electron collision cross sections and radiation chemistry

    International Nuclear Information System (INIS)

    Hatano, Y.

    1983-01-01

    A survey is given of the cross section data needs in radiation chemistry, and of the recent progress in electron impact studies on dissociative excitation of molecules. In the former some of the important target species, processes, and collision energies are presented, while in the latter it is demonstrated that radiation chemistry is a source of new ideas and information in atomic collision research. 37 references, 4 figures

  19. Absolute photoionization cross sections of atomic oxygen

    Science.gov (United States)

    Samson, J. A. R.; Pareek, P. N.

    1985-01-01

    The absolute values of photoionization cross sections of atomic oxygen were measured from the ionization threshold to 120 A. An auto-ionizing resonance belonging to the 2S2P4(4P)3P(3Do, 3So) transition was observed at 479.43 A and another line at 389.97 A. The experimental data is in excellent agreement with rigorous close-coupling calculations that include electron correlations in both the initial and final states.

  20. Total dissociation cross section of halo nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Formanek, J. [Karlova Univ., Prague (Czech Republic). Fakulta Matematicko-Fyzikalni; Lombard, R.J. [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire

    1996-10-01

    Calculations of the total dissociation cross section is performed in the impact parameter representation. The case of {sup 11}Be and {sup 11}Li loosing one and two neutron(s), respectively, by collision on a {sup 12}C target, which remains in its ground state are discussed. The results are found to depend essentially on the rms radius of the halo wave function. (author). 12 refs.

  1. Measurements of Fission Cross Sections of Actinides

    CERN Multimedia

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

    A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.

  2. Cross section of the CMS solenoid

    CERN Multimedia

    Tejinder S. Virdee, CERN

    2005-01-01

    The pictures show a cross section of the CMS solenoid. One can see four layers of the superconducting coil, each of which contains the superconductor (central part, copper coloured - niobium-titanium strands in a copper coating, made into a "Rutherford cable"), surrounded by an ultra-pure aluminium as a magnetic stabilizer, then an aluminium alloy as a mechanical stabilizer. Besides the four layers there is an aluminium mechanical piece that includes pipes that transport the liquid helium.

  3. Cross sections for multistep direct reactions

    International Nuclear Information System (INIS)

    Demetriou, Paraskevi; Marcinkowski, Andrzej; Marianski, Bohdan

    2002-01-01

    Inelastic scattering and charge-exchange reactions have been analysed at energies ranging from 14 to 27 MeV using the modified multistep direct reaction theory (MSD) of Feshbach, Kerman and Koonin. The modified theory considers the non-DWBA matrix elements in the MSD cross section formulae and includes both incoherent particle-hole excitations and coherent collective excitations in the continuum, according to the prescriptions. The results show important contributions from multistep processes at all energies considered. (author)

  4. Capture cross sections for very heavy systems

    International Nuclear Information System (INIS)

    Rowley, N.; Grar, N.; Ntshangase, S.S.

    2006-01-01

    In intermediate-mass systems, collective excitations of the target and projectile can greatly enhance the sub-barrier capture cross section σ cap by giving rise to a distribution of Coulomb barriers. For such systems, capture essentially leads directly to fusion (formation of a compound nucleus (CN)), which then decays through the emission of light particles (neutrons, protons, and alpha particles). Thus the evaporation-residue (ER) cross section is essentially equal to σ cap . For heavier systems the experimental situation is significantly more complicated due to the presence of quasifission (QF) (rapid separation into two fragments before the CN is formed) and by fusion-fission (FF) of the CN itself. Thus three cross sections need to be measured in order to evaluate σ cap . Although the ER essentially recoil along the beam direction. QF and FF fragments are scattered to all angles and require the measurement of angular distribution in order to obtain the excitation function and barrier distribution for capture. Two other approaches to this problem exist. If QF is not important, one can still measure just the ER cross section and try to reconstruct the corresponding σ cap through use of an evaporation-model code that takes account of the FF degree of freedom. Some earlier results on σ cap obtained in this way will be re-analyzed with detail coupled-channels calculations, and the extra-push phenomenon discussed. One may also try to obtain σ cap by exploiting unitarity, that is, by measuring instead the flux of particles corresponding to quasielastic (QE) scattering from the Coulomb barrier. Some new QE results obtained for the 86 Kr + 208 Pb system at iThemba LABS in South Africa will be presented [ru

  5. Inclusive jet cross section at D0

    International Nuclear Information System (INIS)

    Bhattacharjee, M.

    1996-09-01

    Preliminary measurement of the central (|η| ≤ 0.5) inclusive jet cross sections for jet cone sizes of 1.0, 0.7, and 0.5 at D null based on the 1992-1993 (13.7 pb -1 ) and 1994-1995 (90 pb -1 ) data samples are presented. Comparisons to Next-to-Leading Order (NLO) Quantum Chromodynamics (QCD) calculations are made

  6. Fully double-logarithm-resummed cross sections

    International Nuclear Information System (INIS)

    Albino, S.; Bolzoni, P.; Kniehl, B.A.; Kotikov, A.

    2011-01-01

    We calculate the complete double logarithmic contribution to cross sections for semi-inclusive hadron production in the modified minimal-subtraction (MS-bar) scheme by applying dimensional regularization to the double logarithm approximation. The full double logarithmic contribution to the coefficient functions for inclusive hadron production in e + e - annihilation is obtained in this scheme for the first time. Our result agrees with all fixed order results in the literature, which extend to next-to-next-to-leading order.

  7. Atomic-process cross section data, 1

    International Nuclear Information System (INIS)

    1974-12-01

    Compiled by the Data Study Group, the data are intended for fusion plasma physics research. Cross sections of the latest experimental and theoretic studies cover the processes involving H,D,T as principal plasma materials as well as photons and electrons: emission and absorption of electromagnetic wave, electron collision, ion collision, recombination, neutral atom mutual collision, etc. Edition is so made to enable the future renewal by users. (J.P.N.)

  8. Measurement of the fission cross-section of {sup 235}U and {sup 239}Pu for thermal neutrons; Mesures des sections de fission de {sup 235}U et de {sup 239}Pu en neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Fraysse, G; Prosdocimi, A; Netter, F; Samour, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Improved techniques of fast detection have been applied for determining the fission cross-sections of {sup 235}U and {sup 239}Pu with reference to the absorption cross-section of Boron. Monochromatic neutron beams of 0.0322 eV, 0.0626 eV and 0.275 eV have been employed. Use has been made of a Xe-filled gaseous scintillator and of a low-geometry solid state ion chamber. Both measured alpha and fission rates. The results at the reference energy of 0.0253 eV are: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (authors) [French] Des techniques avancees de comptage rapide ont ete mise en oeuvre pour determiner la section efficace de fission de {sup 235}U et de {sup 239}Pu par rapport a celle d'absorption du bore. Des faisceaux de neutrons monochromatiques de 0,0322 eV, 0,0626 eV et 0,275 eV ont ete employes. Les detecteurs utilises sont un scintillateur gazeux rempli de xenon et une chambre d'ionisation a etat solide a basse geometrie. Les deux ont mesure les taux des desintegrations alpha et des fissions. Les resultats a l'energie de reference de 0,0253 eV sont: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (auteurs)

  9. AFCI-2.0 Neutron Cross Section Covariance Library

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

    2011-03-01

    The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

  10. AFCI-2.0 Neutron Cross Section Covariance Library

    International Nuclear Information System (INIS)

    Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

    2011-01-01

    The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

  11. Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

    International Nuclear Information System (INIS)

    Ganesan, S.

    1988-01-01

    This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

  12. Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-04-15

    Zirconium and zircaloy-2 alloy, as constructive materials, have found wide application in reactor technology, especially in heavy water systems for two reasons: a) low neutron absorption cross section, b) good mechanical properties. The thickness of the zirconium and zircaloy-2 for different applications varies from several tenths of a millimeter to about ten millimeters. Therefore, to calculate reactor systems it is desirable to know the effective neutron absorption cross section for the range of thicknesses mention above. The thermal neutron cross sections for these materials are low and no appreciable variation of the effective neutron cross section occurs even for the largest thicknesses. However, this is not true for effective resonance absorption. On the other hand, due to the lack of detailed knowledge of the zirconium resonances, calculations of the effective resonance integrals cannot be performed. Therefore it is necessary to measure the effective total and resonance absorption cross section for zirconium (author)

  13. New techniques for multi-level cross section calculation and fitting

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1981-01-01

    A number of recent developments in multi-level cross section work are described. A new iteration scheme for the conversion of Reich-Moore resonance parameters to Kapur-Peierls parameters allows application of Turing's method for Gaussian broadening of meromorphic functions directly to multi-level cross section expressions, without recourse to the Voigt profiles psi and chi. This makes calculation of Doppler-broadened Reich-Moore and MLBW cross sections practically as fast as SLBW and Adler-Adler cross section calculations involving the Voigt profiles. A convenient distant-level treatment utilizing average resonance parameters is presented. Apart from effectively dealing with edge effects in resonance fitting work it also leads to a simple prescription for the determination of bound levels which reproduce the thermal cross sections correctly. A brief discussion of improved resonance shape fitting techniques is included, with emphasis on the importance of correlated errors and proper use of prior information by application of Bayes' theorem

  14. Measurements of effective total macroscopic cross sections and effective energy of continuum beam

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Hisao [Rikkyo Univ., Yokosuka, Kanagawa (Japan). Inst. for Atomic Energy

    1998-03-01

    Two practically useful quantities are introduced in this study to characterize a continuum neutron beam and to describe transmission phenomena of the beam in field of quantitative neutron radiography: an effective energy instead of a peak energy or a mean energy of the spectrum and an effective total macroscopic (ETM) cross section instead of a total macroscopic (TM) cross section defined at the monochromatic energy. The effective energy was evaluated by means of energy dependence of ETM cross section. To realize the method a beam quality indicator (BQI) has been proposed recently. Several effective energies were measured for non-filtered, filtered neutron beams, and outputs of neutron guide tubes in world by the BQI. A thermal neutron beam and three beams modulated by Pb filters with different thicknesses are studied to measure ETM cross sections for various materials and summarized in a table. Validity of the effective energy determined by the BQI is discussed relating with ETM cross sections of materials. (author)

  15. Neutron capture cross section of ^243Am

    Science.gov (United States)

    Jandel, M.

    2009-10-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL) was used for neutron capture cross section measurement on ^243Am. The high granularity of DANCE (160 BaF2 detectors in a 4π geometry) enables the efficient detection of prompt gamma-rays following neutron capture. DANCE is located on the 20.26 m neutron flight path 14 (FP14) at the Manuel Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE). The methods and techniques established in [1] were used for the determination of the ^243Am neutron capture cross section. The cross sections were obtained in the range of neutron energies from 0.02 eV to 400 keV. The resonance region was analyzed using SAMMY7 and resonance parameters were extracted. The results will be compared to existing evaluations and calculations. Work was performed under the auspices of the U.S. Department of Energy at Los Alamos National Laboratory by the Los Alamos National Security, LLC under Contract No. DE-AC52-06NA25396 and at Lawrence Livermore National Laboratory by the Lawrence Livermore National Security, LLC under Contract No. DE-AC52-07NA27344. [4pt] [1] M. Jandel et al., Phys. Rev. C78, 034609 (2008)

  16. MXS cross-section preprocessor user's manual

    International Nuclear Information System (INIS)

    Parker, F.; Ishikawa, M.; Luck, L.

    1987-03-01

    The MXS preprocessor has been designed to reduce the execution time of programs using isotopic cross-section data and to both reduce the execution time and improve the accuracy of shielding-factor interpolation in the SIMMER-II accident analysis program. MXS is a dual-purpose preprocessing code to: (1) mix isotopes into materials and (2) fit analytic functions to the shelf-shielding data. The program uses the isotope microscopic neutron cross-section data from the CCCC standard interface file ISOTXS and the isotope Bondarenko self-shielding data from the CCCC standard interface file BRKOXS to generate cross-section and self-shielding data for materials. The materials may be a mixture of several isotopes. The self-shielding data for the materials may be the actual shielding factors or a set of coefficients for functions representing the background dependence of the shielding factors. A set of additional data is given to describe the functions necessary to interpolate the shielding factors over temperature

  17. Neutron capture cross sections of Kr

    Directory of Open Access Journals (Sweden)

    Fiebiger Stefan

    2017-01-01

    Full Text Available Neutron capture and β− -decay are competing branches of the s-process nucleosynthesis path at 85Kr [1], which makes it an important branching point. The knowledge of its neutron capture cross section is therefore essential to constrain stellar models of nucleosynthesis. Despite its importance for different fields, no direct measurement of the cross section of 85Kr in the keV-regime has been performed. The currently reported uncertainties are still in the order of 50% [2, 3]. Neutron capture cross section measurements on a 4% enriched 85Kr gas enclosed in a stainless steel cylinder were performed at Los Alamos National Laboratory (LANL using the Detector for Advanced Neutron Capture Experiments (DANCE. 85Kr is radioactive isotope with a half life of 10.8 years. As this was a low-enrichment sample, the main contaminants, the stable krypton isotopes 83Kr and 86Kr, were also investigated. The material was highly enriched and contained in pressurized stainless steel spheres.

  18. Quality Quantification of Evaluated Cross Section Covariances

    International Nuclear Information System (INIS)

    Varet, S.; Dossantos-Uzarralde, P.; Vayatis, N.

    2015-01-01

    Presently, several methods are used to estimate the covariance matrix of evaluated nuclear cross sections. Because the resulting covariance matrices can be different according to the method used and according to the assumptions of the method, we propose a general and objective approach to quantify the quality of the covariance estimation for evaluated cross sections. The first step consists in defining an objective criterion. The second step is computation of the criterion. In this paper the Kullback-Leibler distance is proposed for the quality quantification of a covariance matrix estimation and its inverse. It is based on the distance to the true covariance matrix. A method based on the bootstrap is presented for the estimation of this criterion, which can be applied with most methods for covariance matrix estimation and without the knowledge of the true covariance matrix. The full approach is illustrated on the 85 Rb nucleus evaluations and the results are then used for a discussion on scoring and Monte Carlo approaches for covariance matrix estimation of the cross section evaluations

  19. NNLO jet cross sections by subtraction

    Science.gov (United States)

    Somogyi, G.; Bolzoni, P.; Trócsányi, Z.

    2010-08-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 06, 024 (2005), arXiv:hep-ph/0502226; G. Somogyi and Z. Trócsányi, (2006), arXiv:hep-ph/0609041; G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 01, 070 (2007), arXiv:hep-ph/0609042; G. Somogyi and Z. Trócsányi, JHEP 01, 052 (2007), arXiv:hep-ph/0609043] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  20. 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

  1. Re/Os cosmochronometer: measurement of neutron cross sections

    International Nuclear Information System (INIS)

    Mosconi, M.

    2007-01-01

    This experimental work is devoted to the improved assessment of the Re/Os cosmochronometer. The dating technique is based on the decay of 187 Re (t 1/2 =41.2 Gyr) into 187 Os and determines the age of the universe by the time of onset of nucleosynthesis. The nucleosynthesis mechanisms, which are responsible for the 187 Re/ 187 Os pair, provide the possibility to identify the radiogenic fraction of 187 Os exclusively by nuclear physics considerations. Apart from its radiogenic component, 187 Os can be synthesized otherwise only by the s process, which means that this missing fraction can be reliably determined and subtracted by proper s-process modeling. On the other hand, 187 Re is almost completely produced by the r process. The only information needed for the interpretation as a cosmic clock is the production rate of 187 Re as a function of time. The accuracy of the s-process calculations that are needed to determine the nucleosynthetic abundance of 187 Os depends on the quality of the neutron capture cross sections averaged over the thermal neutron spectrum at the s-process sites. Laboratory measurements of these cross sections have to be corrected for the effect of nuclear levels, which can be significantly populated at the high stellar temperatures during the s process. The neutron capture cross sections of 186 Os, 187 Os and 188 Os have been measured at the CERN n TOF facility in the range between 0.7 eV and 1 MeV. From these data, Maxwellian averaged cross sections have been determined for thermal energies from 5 to 100 keV with an accuracy around 4%, 3%, and 5% for 186 Os, 187 Os, and 188 Os, respectively. Since, the first excited state in 187 Os occurs at 9.75 keV, the cross section of this isotope requires a substantial correction for thermal population of low lying nuclear levels. This effect has been evaluated on the basis of resonance data derived in the (n, γ) experiments and by an improved measurements of the inelastic scattering cross section for

  2. Re/Os cosmochronometer: measurement of neutron cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Mosconi, M.

    2007-12-21

    This experimental work is devoted to the improved assessment of the Re/Os cosmochronometer. The dating technique is based on the decay of {sup 187}Re (t{sub 1/2}=41.2 Gyr) into {sup 187}Os and determines the age of the universe by the time of onset of nucleosynthesis. The nucleosynthesis mechanisms, which are responsible for the {sup 187}Re/{sup 187}Os pair, provide the possibility to identify the radiogenic fraction of {sup 187}Os exclusively by nuclear physics considerations. Apart from its radiogenic component, {sup 187}Os can be synthesized otherwise only by the s process, which means that this missing fraction can be reliably determined and subtracted by proper s-process modeling. On the other hand, {sup 187}Re is almost completely produced by the r process. The only information needed for the interpretation as a cosmic clock is the production rate of {sup 187}Re as a function of time. The accuracy of the s-process calculations that are needed to determine the nucleosynthetic abundance of {sup 187}Os depends on the quality of the neutron capture cross sections averaged over the thermal neutron spectrum at the s-process sites. Laboratory measurements of these cross sections have to be corrected for the effect of nuclear levels, which can be significantly populated at the high stellar temperatures during the s process. The neutron capture cross sections of {sup 186}Os, {sup 187}Os and {sup 188}Os have been measured at the CERN n TOF facility in the range between 0.7 eV and 1 MeV. From these data, Maxwellian averaged cross sections have been determined for thermal energies from 5 to 100 keV with an accuracy around 4%, 3%, and 5% for {sup 186}Os, {sup 187}Os, and {sup 188}Os, respectively. Since, the first excited state in {sup 187}Os occurs at 9.75 keV, the cross section of this isotope requires a substantial correction for thermal population of low lying nuclear levels. This effect has been evaluated on the basis of resonance data derived in the (n, {gamma

  3. Neutron cross-section determination in geological samples (U)

    International Nuclear Information System (INIS)

    Harris, J.M.; McDaniel, P.J.

    1982-01-01

    The Prompt Gamma Neutron Activation Analysis (PGAA) technique yields elemental composition data which can be used to calculate the macroscopic cross section for any sample. The Small Sample Reactivity Measurements (SSRM) technique yields the macroscopic thermal absorption directly. Experimentally, PGAA is somewhat more difficult because of the calibration and data handling than is SSRM. However, SSRM requires a mathematical model of the reactor which means a rather complicated analysis. Once the model and calibration are completed, data analysis is routine. The SSRM technique is production oriented. 9 figures

  4. Illusion thermodynamics: A camouflage technique changing an object into another one with arbitrary cross section

    International Nuclear Information System (INIS)

    He, Xiao; Wu, Linzhi

    2014-01-01

    The previously reported magical thermal devices, such as the thermal invisible cloak and the thermal concentrator, are generalized into one general case named here thermal illusion device. The thermal illusion device is displayed by the design of a thermal reshaper which can reshape an arbitrary thermal object into another one with arbitrary cross section. General expressions of the material parameters for the thermal reshaper are derived unambiguously to greatly facilitate the design of general thermal illusion device. We believe that this work will broaden the current research and pave a path to the thermal invisibility. Numerical simulations show good agreement with the analytical results of the thermal illusion device

  5. Illusion thermodynamics: A camouflage technique changing an object into another one with arbitrary cross section

    Energy Technology Data Exchange (ETDEWEB)

    He, Xiao; Wu, Linzhi, E-mail: wlz@hit.edu.cn [Center for Composite Materials, Harbin Institute of Technology, Harbin 150001 (China)

    2014-12-01

    The previously reported magical thermal devices, such as the thermal invisible cloak and the thermal concentrator, are generalized into one general case named here thermal illusion device. The thermal illusion device is displayed by the design of a thermal reshaper which can reshape an arbitrary thermal object into another one with arbitrary cross section. General expressions of the material parameters for the thermal reshaper are derived unambiguously to greatly facilitate the design of general thermal illusion device. We believe that this work will broaden the current research and pave a path to the thermal invisibility. Numerical simulations show good agreement with the analytical results of the thermal illusion device.

  6. Measurement of thermal neutron cross-sections and resonance integrals for sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As by using sup 2 sup 4 sup 1 Am-Be isotopic neutron source

    CERN Document Server

    Karadag, M; Tan, M; Oezmen, A

    2003-01-01

    Thermal neutron cross-sections and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions were measured by the activation method. The experimental samples with and without a cylindrical Cd shield case in 1 mm wall thickness were irradiated in an isotropic neutron field of the sup 2 sup 4 sup 1 Am-Be neutron source. The induced activities in the samples were measured by high-resolution gamma-ray spectrometry with a calibrated reverse-electrode germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions have been obtained relative to the reference values, sigma sub 0 =13.3+-0.1 b and I sub 0 =14.0+-0.3 b for the sup 5 sup 5 Mn(n,gamma) sup 5 sup 6 Mn reaction as a single comparator. The necessary correction factors for gamma attenuation, thermal neutron and resonance neutron self-shielding effects were taken into...

  7. Measurements of the neutron capture cross sections and incineration potentials of minor-actinides in high thermal neutron fluxes: Impact on the transmutation of nuclear wastes; Mesures des sections efficaces de capture et potentiels d'incineration des actinides mineurs dans les hauts flux de neutrons: Impact sur la transmutation des dechets

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O

    2007-10-15

    This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutron fluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

  8. Neutron-induced fission cross sections

    International Nuclear Information System (INIS)

    Weigmann, H.

    1991-01-01

    In the history of fission research, neutron-induced fission has always played the most important role. The practical importance of neutron-induced fission rests upon the fact that additional neutrons are produced in the fission process, and thus a chain reaction becomes possible. The practical applications of neutron-induced fission will not be discussed in this chapter, but only the physical properties of one of its characteristics, namely (n,f) cross sections. The most important early summaries on the subject are the monograph edited by Michaudon which also deals with the practical applications, the earlier review article on fission by Michaudon, and the review by Bjornholm and Lynn, in which neutron-induced fission receives major attention. This chapter will attempt to go an intermediate way between the very detailed theoretical treatment in the latter review and the cited monograph which emphasizes the applied aspects and the techniques of fission cross-section measurements. The more recent investigations in the field will be included. Section II will survey the properties of cross sections for neutron-induced fission and also address some special aspects of the experimental methods applied in their measurement. Section Ill will deal with the formal theory of neutron-induced nuclear reactions for the resolved resonance region and the region of statistical nuclear reactions. In Section IV, the fission width, or fission transmission coefficient, will be discussed in detail. Section V will deal with the broader structures due to incompletely damped vibrational resonances, and in particular will address the special case of thorium and neighboring isotopes. Finally, Section VI will briefly discuss parity violation effects in neutron-induced fission. 74 refs., 14 figs., 3 tabs

  9. Pile oscillator measurements of thermal absorption cross sections of Al, Mg, Fe and Cu; Mesure des sections efficaces thermiques d'absorption de l'Al, du Mg, du Fe et du Cu par la methode d'oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Carre, J C; Vidal, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The phase oscillation technique used at ZOE has the property of reducing of a marked factor the effect of neutron scattering by the sample. The absorption cross sections of poorly absorbing and highly scattering materials have been measured; for neutrons of 2,200 m/s, the following values are obtained: 229 {+-} 3 mb for Al; 64.2 {+-} 1.5 mb for Mg, 2.53 {+-} 0.03 b for Fe and 3.74 {+-} 0.04 b for Cu. (authors) [French] La methode d'oscillation de phase utilisee a ZOE possede la propriete de diminuer d'un facteur important l'influence de la diffusion. Les sections efficaces de capture de materiaux peu capturants et tres diffusants, ont ete mesurees; pour des neutrons de 2200 m/s, on obtient pour l'Al: 229 {+-} 3mb, pour le Mg: 64,2 {+-} 1,5 mb, pour le Fe: 2,53 {+-} 0,03 b et pour le Cu: 3,74 {+-} 0,04 b. (auteurs)

  10. LEP vacuum chamber, cross-section

    CERN Multimedia

    CERN PhotoLab

    1983-01-01

    Cross-section of the final prototype for the LEP vacuum chamber. The elliptic main-opening is for the beam. The small channel to the left is for the cooling water, to carry away the heat deposited by the synchrotron radiation. The square channel to the right houses the Non-Evaporable Getter (NEG) pump. The chamber is made from extruded aluminium. Its outside is clad with lead, to stop the synchrotron radiation emitted by the beam. For good adherence between Pb and Al, the Al chamber was coated with a thin layer of Ni. Ni being slightly magnetic, some resulting problems had to be overcome. See also 8301153.

  11. How to extract cross sections from TDHF

    International Nuclear Information System (INIS)

    Le Tourneux, Jean

    1979-01-01

    In spite of all the recent progress in solving numerically TDHF (Time Dependent Hartree-Fock) equations for heavy-ion collisions, this method is still far from lending itself readily to the computation of cross sections, except in the case of fusion. The theory presented here is purely formal so far and would lead to fairly heavy calculations in practice. It solves the problem of channel identification in the outgoing asymptotic region of TDHF solutions. It throws a bridge between TDHF and more traditional theories of nuclear reactions, which are time-independent

  12. Hyperon magnetic moments and total cross sections

    International Nuclear Information System (INIS)

    Lipkin, H.J.

    1982-06-01

    The new data on both total cross sections and magnetic moments are simply described by beginning with the additive quark model in an SU(3) limit where all quarks behave like strange quarks and breaking both additivity and SU(3) simultaneously with an additional non-additive mechanism which affects only nonstrange quark contributions. The suggestion that strange quarks behave more simply than nonstrange may provide clues to underlying structure or dynamics. Small discrepancies in the moments are analyzed and shown to provide serious difficulties for most models if they are statistically significant. (author)

  13. Neutron capture cross section of $^{93}$Zr

    CERN Document Server

    We propose to measure the neutron capture cross section of the radioactive isotope $^{93}$Zr. This project aims at the substantial improvement of existing results for applications in nuclear astrophysics and emerging nuclear technologies. In particular, the superior quality of the data that can be obtained at n_TOF will allow on one side a better characterization of s-process nucleosynthesis and on the other side a more accurate material balance in systems for transmutation of nuclear waste, given that this radioactive isotope is widely present in fission products.

  14. Fission cross section measurements for minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    Fursov, B. [IPPE, Obninsk (Russian Federation)

    1997-03-01

    The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)

  15. Charge changing cross sections of relativistic uranium

    International Nuclear Information System (INIS)

    Gould, H.; Greiner, D.; Lindstrom, P.; Symons, T.J.M.; Crawford, H.; Thieberger, P.; Wegner, H.

    1984-11-01

    We report equilibrium charge state distributions of uranium at energies of 962 MeV/nucleon, 437 MeV/nucleon and 200 MeV/nucleon in low Z and high Z targets and the cross sections for U 92+ reversible U 91+ and U 91+ reversible U 90+ at 962 MeV/nucleon and 437 MeV/nucleon. Equilibrium thickness Cu targets produce approx. = 5% bare U 92+ at 200 MeV/nucleon and 85% U 92+ at 962 MeV/nucleon. 7 references, 5 figures

  16. Charge changing cross sections of relativistic uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gould, H; Greiner, D; Lindstrom, P; Symons, T J.M.; Crawford, H; Thieberger, P; Wegner, H

    1985-05-15

    We report equilibrium charge state distributions of uranium at energies of 962 MeV/nucleon, 437 MeV/nucleon and 200 MeV/nucleon in low Z and high Z targets and the cross sections for U/sup 92 +/reversibleU/sup 91 +/ and U/sup 91 +/reversibleU/sup 90 +/ at 962 MeV/nucleon and 437 MeV/nucleon. Equilibrium thickness Cu targets produce approx.=5% bare U/sup 92 +/ at 200 MeV/nucleon and 85% U/sup 92 +/ at 962 MeV/nucleon.

  17. Empirical continuation of the differential cross section

    International Nuclear Information System (INIS)

    Borbely, I.

    1978-12-01

    The theoretical basis as well as the practical methods of empirical continuation of the differential cross section into the nonphysical region of the cos theta variable are discussed. The equivalence of the different methods is proved. A physical applicability condition is given and the published applications are reviewed. In many cases the correctly applied procedure turns out to provide nonsignificant or even incorrect structure information which points to the necessity for careful and statistically complete analysis of the experimental data with a physical understanding of the analysed process. (author)

  18. L-shell photoelectric cross section measurements

    Energy Technology Data Exchange (ETDEWEB)

    Arora, S K; Allawadhi, K L; Sood, B S [Punjabi Univ., Patiala (India). Nuclear Science Labs.

    1981-05-14

    L-shell photoelectric cross sections in Ta, W, Au, Pb, Th and U at 59.5 keV have been determined using three different versions of Sood's method of measuring the absolute yield of fluorescent x-rays when a target is irradiated with a known flux of photons. The results obtained by all the methods agree with one another showing that no hidden systematic errors are involved in the measurements. The present results are found to compare well with the theoretical calculations of Scofield (Lawrence Livermore Laboratory Report No 51326).

  19. Total cross section measurement of radioactive isotopes with a thin beam neutron spectrometer

    International Nuclear Information System (INIS)

    Razbudej, V.F.; Vertebnyj, V.P.; Padun, G.S.; Muravitskij, A.V.

    1975-01-01

    The method for measuring the neutron total cross sections of radioactive isotopes by a time-of-flight spectrometer with a narrow (0.17 mm in diameter) beam of thermal neutrons is described. The distinguishing feature of this method is the use of capillary samples with a small amount of substance (0.05-1.0 mg). The energy range is 0.01-0.3 eV. The total cross sections of irradiated samples of sub(153)Eu and sub(151)Eu are measured. From them are obtained the cross sections of sub(152)Eu (Tsub(1/2)=12.4 g) and of sub(154)E (Tsub(1/2)=8.6 yr); they equal 11400+-1400 and 1530+-190 barn at E=0.0253 eV. The cross section of the sub(152)Eu absorption for the thermal spectrum (T=333 K) is determined by the activation method; it is 8900+-1200 barn

  20. Fission cross section measurements at the LLL 100-MeV linac

    International Nuclear Information System (INIS)

    Browne, J.C.

    1975-01-01

    The fission cross section for 235 U was measured from thermal energy to 20 MeV in several steps. First, the cross section was measured from 8 MeV to 20 MeV relative to the n,p scattering cross section and then from thermal to one MeV relative to 6 Li(n,α). In addition, a measurement of the ratio of the fission cross sections of 235 U and 238 U relative to 235 U has been completed in the range 1 keV to 30 MeV for 233 U and 100 keV to 30 MeV for 238 U. Statistical uncertainties are less than 4 percent. (U.S.)

  1. Impact of newly-measured gadolinium cross sections on BWR fuel rod reaction rate distributions

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.; Perret, G.; Murphy, M.; Grimm, P.; Seiler, R. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federal de Lausanne, CH-1015 Lausanne (Switzerland)

    2008-07-01

    Recent measurements of capture and total cross sections performed at the Rensselaer Polytechnic Institute in the USA confirmed many of the gadolinium thermal and resonant neutron cross section parameters within uncertainties, but they also showed up important discrepancies well out of uncertainties, such as an approx11% overestimation of the {sup 157}Gd thermal capture cross section in ENDF/B-VI and -VII with respect to the newly measured data. In this work, the impact of the newly measured gadolinium cross sections on BWR reactor physics parameters has been preliminarily evaluated. The comparisons of rod-by-rod fission rate and modified conversion ratio predictions with selected cold critical experiments at the PROTEUS reactor in Switzerland show the potential to resolve long-term unexplained discrepancies. (authors)

  2. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  3. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

    2005-01-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  4. Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

    International Nuclear Information System (INIS)

    Kasselmann, S.; Druska, C.; Lauer, A.

    2010-01-01

    The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

  5. Solution of multi-group diffusion equation in x-y-z geometry by finite Fourier transformation

    International Nuclear Information System (INIS)

    Kobayashi, Keisuke

    1975-01-01

    The multi-group diffusion equation in three-dimensional x-y-z geometry is solved by finite Fourier transformation. Applying the Fourier transformation to a finite region with constant nuclear cross sections, the fluxes and currents at the material boundaries are obtained in terms of the Fourier series. Truncating the series after the first term, and assuming that the source term is piecewise linear within each mesh box, a set of coupled equations is obtained in the form of three-point equations for each coordinate. These equations can be easily solved by the alternative direction implicit method. Thus a practical procedure is established that could be applied to replace the currently used difference equation. This equation is used to solve the multi-group diffusion equation by means of the source iteration method; and sample calculations for thermal and fast reactors show that the present method yields accurate results with a smaller number of mesh points than the usual finite difference equations. (auth.)

  6. Electron capture cross-section of Au-Fe complex in silicon

    International Nuclear Information System (INIS)

    Ali, Akbar; Shafi, M; Majid, Abdul

    2006-01-01

    A deep level transient spectroscopy technique is applied to study the capture cross-section of an iron-gold complex. The thermal ionization energy obtained from emission rate data is found to be E c -0.36 eV. The Au-Fe complex is a single defect having a capture cross-section of 2.48x10 -16 cm 2 for electrons which is independent of temperature

  7. The l-mixing cross section of Rydberg states of atomic Rb and the scaling LAW

    International Nuclear Information System (INIS)

    Liu Hong; Chen Aiqiu; Li Baiwen

    1991-01-01

    On the basis of impulse approximate method, a kind of analytical wavefunctions based on a potential model was used to calculate the l mixing cross section of thermal collision of Rydberg states of atomic Rb with rare gas (He, Ne). The results were compared with the experimental results and other theoretical values. These results show that there exists a kind of scaling law for the l mixing cross section of Rydberg alkali atoms

  8. C4P cross-section libraries for safety analyses with SIMMER and related studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

    2011-01-01

    A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

  9. Electron capture cross-section of Au-Fe complex in silicon

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Akbar; Shafi, M; Majid, Abdul [Advance Materials Physics Laboratory, Department of Physics, Quaid-i-Azam University, Islamabad (Pakistan)

    2006-10-15

    A deep level transient spectroscopy technique is applied to study the capture cross-section of an iron-gold complex. The thermal ionization energy obtained from emission rate data is found to be E{sub c} -0.36 eV. The Au-Fe complex is a single defect having a capture cross-section of 2.48x10{sup -16} cm{sup 2} for electrons which is independent of temperature.

  10. ZZ SNLRML, Dosimetry Cross-Section Recommendations

    International Nuclear Information System (INIS)

    1996-01-01

    Description of program or function: Format: SAND-II; Number of groups: 640 group SAND-II group structure. Nuclides: Cd, B, Au, S, Ni, Li, F, Na, Mg, Al, Si, P, Sc, Ti, Mn, Fe, Co, Cu, Zn, Zr, Nb, Mo, Rh, Ag, In, I, Th, U, Np, Pu, Am. Origin: ENDF/B-VI, ENDF/B-V, IRDF-90, JENDL-3, JEF 2.2 and GLUCS data with special modifications from private communications. Weighting spectrum: flat. SNLRML is a reactor dosimetry library that draws upon all available evaluated cross section libraries and selects the best evaluation for application to research reactor spectrum determinations. Many of the components of the SNLRML come from the ENDF/B-VI and IRDF-90 (DLC-0161) libraries. The library format was selected for easy interface with spectrum determination codes such as SAND-II (CCC-0112 and LSL-M2 (PSR-233) and the new PSR-0345/SNL/SAND-II has been enhanced to interface with SNLRML. The data is recommended for spectrum determination applications and for the prediction of neutron activation of typical radiation sensor materials. The library has been tested for consistency of the cross section in wide variety of neutron environments. The results and cautions from this testing have been documented. The data has been interfaced with radiation transport codes, such as TWODANT-SYS (CCC-0547) and MCNP (CCC-0200), in order to compare calculated and measured activities for benchmark reactor experiments

  11. Pion production cross sections and associated parameters

    International Nuclear Information System (INIS)

    Bradbury, J.N.

    1985-01-01

    Negative pions have been used for radiotherapy at the meson factories LAMPF (USA), SIN (Switzerland), and TRIUMF (Canada) and have been planned for use at new meson facilities under construction (USSR) and at proposed dedicated medical facilities. Providing therapeutically useful dose rates of pions requires a knowledge of the pion production cross sections as a function of primary proton energy (500 to 1000 MeV), pion energy (less than or equal to100 MeV), production angle, and target material. The current status of the data base in this area is presented including theoretical guidelines for extrapolation purposes. The target material and geometry, as well as the proton and pion beam parameters, will affect the electron (and muon) contamination in the beam which may have an important effect on both the LET characteristics of the dose and the dose distribution. In addition to cross-section data, channel characteristics such as length of pion trajectory, solid-angle acceptance, and momentum analysis will affect dose rate, distribution, and quality. Such considerations are briefly addressed in terms of existing facilities and proposed systems. 16 refs., 6 figs

  12. Total neutron cross section for 181Ta

    Directory of Open Access Journals (Sweden)

    Schilling K.-D.

    2010-10-01

    Full Text Available The neutron time of flight facility nELBE, produces fast neutrons in the energy range from 0.1 MeV to 10 MeV by impinging a pulsed relativistic electron beam on a liquid lead circuit [1]. The short beam pulses (∼10 ps and a small radiator volume give an energy resolution better than 1% at 1 MeV using a short flight path of about 6 m, for neutron TOF measurements. The present neutron source provides 2 ⋅ 104  n/cm2s at the target position using an electron charge of 77 pC and 100 kHz pulse repetition rate. This neutron intensity enables to measure neutron total cross section with a 2%–5% statistical uncertainty within a few days. In February 2008, neutron radiator, plastic detector [2] and data acquisition system were tested by measurements of the neutron total cross section for 181Ta and 27Al. Measurement of 181Ta was chosen because lack of high quality data in an anergy region below 700 keV. The total neutron cross – section for 27Al was measured as a control target, since there exists data for 27Al with high resolution and low statistical error [3].

  13. Atlas of photoneutron cross sections obtained with monoenergetic photons

    International Nuclear Information System (INIS)

    Dietrich, S.S.; Berman, B.L.

    1988-01-01

    Photoneutron cross-section and integrated cross-section data obtained with monoenergetic photons are presented in a uniform format. All of the measured partial photoneutron cross sections, the total photoneutron cross section, and the photoneutron yield cross section are plotted as functions of the incident photon energy, as are the integrated photoneutron cross sections and their first and second moments. The values of the integrated cross sections and the moments of the integrated total cross section up to the highest photon energy for which they were measured are tabulated, as are the parameters of Lorentz curves fitted to the total photoneutron cross-section data for medium and heavy nuclei (A>50). This compilation is current as of June 1987. copyright 1988 Academic Press, Inc

  14. Differences between LASL- and ANL-processed cross sections

    International Nuclear Information System (INIS)

    Kidman, R.B.; MacFarlane, R.E.; Becker, M.

    1978-03-01

    As part of the Los Alamos Scientific Laboratory (LASL) cross-section processing development, LASL cross sections and results from MINX/1DX system are compared to the Argonne National Laboratory cross sections and results from the ETOE-2/MC 2 -2 system for a simple reactor problem. Exact perturbation theory is used to establish the eigenvalue effect of every isotope group cross-section difference. Cross sections, cross-section differences, and their eigenvalue effects are clearly and conveniently displayed and compared on a group-by-group basis

  15. Cross section homogenization analysis for a simplified Candu reactor

    International Nuclear Information System (INIS)

    Pounders, Justin; Rahnema, Farzad; Mosher, Scott; Serghiuta, Dumitru; Turinsky, Paul; Sarsour, Hisham

    2008-01-01

    The effect of using zero current (infinite medium) boundary conditions to generate bundle homogenized cross sections for a stylized half-core Candu reactor problem is examined. Homogenized cross section from infinite medium lattice calculations are compared with cross sections homogenized using the exact flux from the reference core environment. The impact of these cross section differences is quantified by generating nodal diffusion theory solutions with both sets of cross sections. It is shown that the infinite medium spatial approximation is not negligible, and that ignoring the impact of the heterogeneous core environment on cross section homogenization leads to increased errors, particularly near control elements and the core periphery. (authors)

  16. Fission neutron spectrum averaged cross sections for threshold reactions on arsenic

    International Nuclear Information System (INIS)

    Dorval, E.L.; Arribere, M.A.; Kestelman, A.J.; Comision Nacional de Energia Atomica, Cuyo Nacional Univ., Bariloche; Ribeiro Guevara, S.; Cohen, I.M.; Ohaco, R.A.; Segovia, M.S.; Yunes, A.N.; Arrondo, M.; Comision Nacional de Energia Atomica, Buenos Aires

    2006-01-01

    We have measured the cross sections, averaged over a 235 U fission neutron spectrum, for the two high threshold reactions: 75 As(n,p) 75 mGe and 75 As(n,2n) 74 As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27 Al(n,p) 27 Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58 Ni(n,p) 58m+g Co averaged cross section, respectively. The measured averaged cross sections were also evaluated semi-empirically by numerically integrating experimental differential cross section data extracted for both reactions from the current literature. The calculations were performed for four different representations of the thermal-neutron-induced 235 U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties. (author)

  17. Creation of problem-dependent Doppler-broadened cross sections in the KENO Monte Carlo code

    International Nuclear Information System (INIS)

    Hart, Shane W.D.; Celik, Cihangir; Maldonado, G. Ivan; Leal, Luiz

    2016-01-01

    Highlights: • A quick method of Doppler broadening one- and two-dimensional cross sections has been added to KENO. • The method uses a finite difference method to broaden data to user defined temperatures. • Various problems and benchmarks were run to showcase results. • Results with the Doppler broadened cross sections are closer to benchmark results. - Abstract: This paper introduces a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. The problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.

  18. MOX Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2003-01-01

    The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

  19. K sup + nucleus total cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Sawafta, R.

    1990-01-01

    The scattering of K{sup +} mesons from nuclei has attracted considerable interest in the last few years. The K{sup +} holds a very special position as the weakest of all strongly interaction probes. The average cross section is not larger than about 10 mb at lab momenta below 800 MeV/c, corresponding to a mean free path in the nucleus larger than 5 fm. Thus the K{sup +} is capable of probing the entire volume of the nucleus. Single scattering of the K{sup +} with a nucleon in the nucleus dominates the nuclear scattering, and only small and calculable higher order corrections are needed. The nucleon is a dynamical entity and its internal structure can, in principle, be altered by its surrounding nuclear environment. This work reports an experiment in which the K{sup +} is used to compare the nucleon in the nucleus with a free nucleon.

  20. Differential cross section of atomic hydrogen photoionization

    International Nuclear Information System (INIS)

    Kondratovich, V.D.; Ostrovskij, V.N.

    1986-01-01

    Differential cross-section of atomic hydrogen photoeffect in external electric field was investigated in semiclassical approximation. Interference was described. It occurred due to the fact that infinite number of photoelectron trajectories leads to any point of classically accessible motion region. Interference picture can reach macroscopic sizes. The picture is determined by location of function nodes, describing finite electron motion along one of parabolic coordinates. The squares of external picture rings are determined only by electric field intensity in the general case at rather high energies. Quantum expression for photocurrent density was obtained using Green function in superposition of Coulomb and uniform field as well as semiclassical approximation. Possible applications of macroscopic interference picture to specification of atom ionization potentials, selective detection of atoms or particular molecules, as well as weak magnetic field and observation of Aaronov-Bom effect are discussed

  1. Angle-averaged Compton cross sections

    International Nuclear Information System (INIS)

    Nickel, G.H.

    1983-01-01

    The scattering of a photon by an individual free electron is characterized by six quantities: α = initial photon energy in units of m 0 c 2 ; α/sub s/ = scattered photon energy in units of m 0 c 2 ; β = initial electron velocity in units of c; phi = angle between photon direction and electron direction in the laboratory frame (LF); theta = polar angle change due to Compton scattering, measured in the electron rest frame (ERF); and tau = azimuthal angle change in the ERF. We present an analytic expression for the average of the Compton cross section over phi, theta, and tau. The lowest order approximation to this equation is reasonably accurate for photons and electrons with energies of many keV

  2. ISSUES IN NEUTRON CROSS SECTION COVARIANCES

    Energy Technology Data Exchange (ETDEWEB)

    Mattoon, C.M.; Oblozinsky,P.

    2010-04-30

    We review neutron cross section covariances in both the resonance and fast neutron regions with the goal to identify existing issues in evaluation methods and their impact on covariances. We also outline ideas for suitable covariance quality assurance procedures.We show that the topic of covariance data remains controversial, the evaluation methodologies are not fully established and covariances produced by different approaches have unacceptable spread. The main controversy is in very low uncertainties generated by rigorous evaluation methods and much larger uncertainties based on simple estimates from experimental data. Since the evaluators tend to trust the former, while the users tend to trust the latter, this controversy has considerable practical implications. Dedicated effort is needed to arrive at covariance evaluation methods that would resolve this issue and produce results accepted internationally both by evaluators and users.

  3. Partial cross sections in H- photodetachment

    International Nuclear Information System (INIS)

    Halka, M.

    1993-04-01

    This dissertation reports experimental measurements of partial decay cross sections in the H - photodetachment spectrum. Observed decays of the 1 P 0 H -** (n) doubly-excitedresonances to the H(N=2) continuum are reported for n=2,3, and 4 from 1990 runs in which the author participated. A recent analysis of 1989 data revealing effects of static electric fields on the partial decay spectrum above 13.5 eV is also presented. The experiments were performed at the High Resolution Atomic Beam Facility. the Los Alamos Meson Physics Facility, with a relativistic H - beam (β=0.842)intersecting a ND:YAG laser. Variation of the intersection angle amounts to Doppler-shifting the photon energy, allowing continuous tuning of the laser energy as viewed from the moving ions' frame

  4. Angle-averaged Compton cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, G.H.

    1983-01-01

    The scattering of a photon by an individual free electron is characterized by six quantities: ..cap alpha.. = initial photon energy in units of m/sub 0/c/sup 2/; ..cap alpha../sub s/ = scattered photon energy in units of m/sub 0/c/sup 2/; ..beta.. = initial electron velocity in units of c; phi = angle between photon direction and electron direction in the laboratory frame (LF); theta = polar angle change due to Compton scattering, measured in the electron rest frame (ERF); and tau = azimuthal angle change in the ERF. We present an analytic expression for the average of the Compton cross section over phi, theta, and tau. The lowest order approximation to this equation is reasonably accurate for photons and electrons with energies of many keV.

  5. Sudakov resummation of multiparton QCD cross sections

    CERN Document Server

    Bonciani, R; Mangano, Michelangelo L; Nason, P

    2003-01-01

    We present the general expressions for the resummation, up to next-to-leading logarithmic accuracy, of Sudakov-type logarithms in processes with an arbirtrary number of hard-scattering partons. These results document the formulae used by the authors in several previous phenomenological studies. The resummation formulae presented here, which are valid for phase-space factorizable observables, determine the resummation correction in a process-independent fashion. All process dependence is encoded in the colour and flavour structure of the leading order and virtual one-loop amplitudes, and in Sudakov weights associated to the cross section kinematics. We explicitly illustrate the application to the case of Drell--Yan and prompt-photon production.

  6. Sudakov resummation of multiparton QCD cross sections

    International Nuclear Information System (INIS)

    Bonciani, Roberto; Catani, Stefano; Mangano, Michelangelo L.; Nason, Paolo

    2003-01-01

    We present the general expressions for the resummation, up to next-to-leading logarithmic accuracy, of Sudakov-type logarithms in processes with an arbitrary number of hard-scattering partons. These results document the formulae used by the authors in several previous phenomenological studies. The resummation formulae presented here, which are valid for phase-space factorizable observables, determine the resummation correction in a process-independent fashion. All process dependence is encoded in the colour and flavour structure of the leading order and virtual one-loop amplitudes, and in Sudakov weights associated to the cross section kinematics. We explicitly illustrate the application to the case of Drell-Yan and prompt-photon production

  7. Electroweak Boson Cross-Section Measurements

    CERN Document Server

    The ATLAS collaboration

    2009-01-01

    This report summarises the ATLAS prospects for the measurement of W and Z pro- duction cross-section at the LHC. The electron and muon decay channels are considered. Focusing on the early data taking phase, strategies are presented that allow a fast and robust extraction of the signals. An overall uncertainty of about 5% can be achieved with 50 pb−1 in the W channels, where the background uncertainty dominates (the luminosity measurement uncertainty is not discussed here). In the Z channels, the expected preci- sion is 3%, the main contribution coming from the lepton selection efficiency uncertainty. Extrapolating to 1 fb−1 , the uncertainties shrink to incompressible values of 1-2%, de- pending on the final state. This irreducible uncertainty is essentially driven by strong interaction effects, notably parton distribution uncertainties and non-perturbative effects, affecting the W and Z rapidity and transverse momentum distributions. These effects can be constrained by measuring these distributions. Al...

  8. Reaction cross section for Ne isotopes

    International Nuclear Information System (INIS)

    Panda, R.N.; Sahu, B.K.; Patra, S.K.

    2012-01-01

    In the present contribution, first the bulk properties are calculated, such as binding energy (BE), root mean square charge radius r ch , matter radius r m and quadrupole deformation parameter β 2 for 18-32 Ne isotopes in the Relativistic mean field (RMF) and effective field theory motivated RMF (E-RMF) formalisms . Then the total nuclear reaction cross section σR is analyzes for the scattering of 20 Ne and 28-32 Ne from a 12 C target at 240 MeV/nucleon by using the RMF model. Thus the objective of the present study is to calculate the bulk properties as well as a systematic analysis of σR over a range of neutron rich nuclei in the frame work of Glauber model

  9. Topological supersymmetric structure of hadron cross sections

    International Nuclear Information System (INIS)

    Gauron, P.; Nicolescu, B.; Ouvry, S.

    1980-12-01

    Recently a way of fully implementing unitarity in the framework of a Dual Topological Unitarization theory, including not only mesons but also baryons, was found. This theory consists in the topological description of hadron interactions involving confined quarks in terms of two 2-dimensional surfaces (a closed 'quantum' surface and a bounded 'classical' surface). We show that this description directly leads, at the zeroth order of the topological expansion, to certain relations between hadron cross-sections, in nice agreement with experimental data. A new topological suppression mechanism is shown to play an important dynamical role. We also point out a new topological supersymmetry property, which leads to realistic experimental consequences. A possible topological origin of the rho and ω universality relations emerges as a by-product of our study

  10. Neutron scattering cross sections of uranium-238

    International Nuclear Information System (INIS)

    Beghian, L.E.; Kegel, G.H.R.; Marcella, T.V.; Barnes, B.K.; Couchell, G.P.; Egan, J.J.; Mittler, A.; Pullen, D.J.; Schier, W.A.

    1979-01-01

    The University of Lowell high-resolution time-of-flight spectrometer was used to measure angular distributions and 90-deg excitation functions for neutrons scattered from 238 U in the energy range from 0.9 to 3.1 MeV. This study was limited to the elastic and the first two inelastic groups, corresponding to states of 238 U at 45 keV (2 + ) and 148 keV (4 + ). Angular distributions were measured at primary neutron energies of 1.1, 1.9, 2.5, and 3.1 MeV for the same three neutron groups. Whereas the elastic data are in fair agreement with the evaluation in the ENDF/B-IV file, there is substantial disagreement between the inelastic measurements and the evaluated cross sections. 12 figures

  11. Electron capture cross sections by O+ from atomic He

    International Nuclear Information System (INIS)

    Joseph, Dwayne C; Saha, Bidhan C

    2009-01-01

    The adiabatic representation is used in both the quantal and semi classical molecular orbital close coupling methods (MOCC) to evaluate charge exchange cross sections. Our results show good agreement with experimental cross sections

  12. Electron capture cross sections by O+ from atomic He

    Science.gov (United States)

    Joseph, Dwayne C.; Saha, Bidhan C.

    2009-11-01

    The adiabatic representation is used in both the quantal and semi classical molecular orbital close coupling methods (MOCC) to evaluate charge exchange cross sections. Our results show good agreement with experimental cross sections

  13. Single-level resonance parameters fit nuclear cross-sections

    Science.gov (United States)

    Drawbaugh, D. W.; Gibson, G.; Miller, M.; Page, S. L.

    1970-01-01

    Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.

  14. Total and ionization cross sections of electron scattering by fluorocarbons

    International Nuclear Information System (INIS)

    Antony, B K; Joshipura, K N; Mason, N J

    2005-01-01

    Electron impact total cross sections (50-2000 eV) and total ionization cross sections (threshold to 2000 eV) are calculated for typical plasma etching molecules CF 4 , C 2 F 4 , C 2 F 6 , C 3 F 8 and CF 3 I and the CF x (x 1-3) radicals. The total elastic and inelastic cross sections are determined in the spherical complex potential formalism. The sum of the two gives the total cross section and the total inelastic cross section is used to calculate the total ionization cross sections. The present total and ionization cross sections are found to be consistent with other theories and experimental measurements, where they exist. Our total cross section results for CF x (x = 1-3) radicals presented here are first estimates on these species

  15. Mid-IR Absorption Cross-Section Measurements of Hydrocarbons

    KAUST Repository

    Alrefae, Majed Abdullah

    2013-01-01

    -known at combustion-relevant conditions. Absorption cross-section is an important spectroscopic quantity and has direct relation to the species concentration. In this work, the absorption cross-sections of basic hydrocarbons are measured using Fourier Transform

  16. Nuclear reactions cross section measurement using Z-pinch technology

    Energy Technology Data Exchange (ETDEWEB)

    Bulgakov, T; Bystritskij, V; Mesyats, G A [Russian Academy of Sciences, Tomsk (Russian Federation). Institute of Electrophysics; and others

    1997-12-31

    Direct experimental estimate of the upper limit of the d + d {yields} {sup 3}He + n cross-section at deuteron energies below the keV region was obtained for the first time. The experiment was performed at the Pulsed Ion Beam Accelerator of the High-Current Electronics Institute in Tomsk, using high intensity, radially converging deuteron beams, generated during implosion of liner plasma. A two-jet liner made of 17% D{sub 2} + 83% N{sub 2} gas was used, with the inner jet serving as the target. The dd-fusion neutrons were registered by time-of-flight scintillator spectrometers and BF{sub 3} detectors of thermal neutrons placed in a polyethylene moderator. The upper limit obtained for the d + d {yields} {sup 3}He + n cross-section for a deuteron energy of 440 eV is {sigma} < 2 x 10{sup -34} cm{sup 2} at the 90% confidence level. The result demonstrates that the liner implosion technique can be used in the investigation of nuclear reactions between light nuclei at infra low energies, previously not accessible in experiments with classical beam accelerators. (author). 7 refs.

  17. Research on Fast-Doppler-Broadening of neutron cross sections

    International Nuclear Information System (INIS)

    Li, S.; Wang, K.; Yu, G.

    2012-01-01

    A Fast-Doppler-Broadening method is developed in this work to broaden Continuous Energy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithm and parallel computing are implemented in this method, which is unprecedented in the history of cross section processing. Compared to the traditional code (NJOY, SIGMA1, etc.), the new Fast-Doppler-Broadening method shows a remarkable speedup with keeping accuracy. The purpose of using Gauss integration is to avoid complex derivation of traditional broadening formula and heavy load of computing complementary error function that slows down the Doppler broadening process. The OpenMP environment is utilized in parallel computing which can take full advantage of modern multi-processor computers. Combination of the two can reduce processing time of main actinides (such as 238 U, 235 U) to an order of magnitude of 1∼2 seconds. This new method is fast enough to be applied to Online Doppler broadening. It can be combined or coupled with Monte Carlo transport code to solve temperature dependent problems and neutronics-thermal hydraulics coupled scheme which is a big challenge for the conventional NJOY-MCNP system. Examples are shown to determine the efficiency and relative errors compared with the NJOY results. A Godiva Benchmark is also used in order to test the ACE libraries produced by the new method. (authors)

  18. Neutron cross section standards and instrumentation: Annual report

    International Nuclear Information System (INIS)

    1987-01-01

    This annual report from the National Bureau of Standards contains a summary of the results of the Neutron Cross Section Standards and Instrumentation Program. The technical measurements for the past year are given along with the proposed program and budget needs for the next three years. The neutron standards measurements have concentrated on the most important 235 U(n,f) cross section in the thermal to 20 MeV energy range along with the development of neutron detectors required for these measurements. The NBS measurements have made a significant contribution to the improvement in the understanding of this reaction. Measurements were performed with numerous neutron detectors at overlapping energies and at different neutron sources in order to reduce the systematic errors to achieve the required accuracy in this important neutron standard. Significant progress was also made in the development of a detector to utilize the 3 He(n,p) reaction as a standard in the eV to MeV energy region. Improvements in data acquisition systems as well as additional studies of advanced neutron sources were accomplished. Contacts with private industry were maintained and coordination of the neutron standards evaluation was continued. The report also includes biographical listings of the research staff along with copies of a few of our recent publications. 13 figs., 1 tab

  19. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

  20. Measurements of Integral Cross Section Ratios in Two Dosimetry Benchmark Neutron Fields

    International Nuclear Information System (INIS)

    Fabry, A.; Czock, K.H.

    1974-12-01

    In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the 103 Rh(n,n') 103m Rh and 58 Ni(n,p) 58 Co integral cross sections have been accurately measured relatively to the 115 In(n,n') 115m In cross section in the 235 U thermal dission neutron spectrum and in the MOLΣΣ Intermediate-Energy Standard Neutron field. In this last neutron field, the data are related also to the 235 U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific 103 Rh(n,n') 103m Rh differential-energy cross section among the existing, conflicting data. (author)

  1. Measurements of integral cross section ratios in two dosimetry benchmark neutron fields

    International Nuclear Information System (INIS)

    Fabry, A.; Czock, K.H.

    1974-12-01

    In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the 103 Rh(n,n') 103m Rh and 58 Ni(n,p) 58 Co integral cross sections have been accurately measured relatively to the 115 In(n,n') 115m In cross section in the 235 U thermal fission neutron spectrum and in the MOL-ΣΣ intermediate-energy standard neutron field. In this last neutron field, the data are related also to the 235 U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific 103 Rh(n,n') 103m Rh differential-energy cross section among the existing, conflicting data. (author)

  2. Measurements of integral cross section ratios in two dosimetry benchmark neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A [CEN-SCK, Mol (Belgium); Czock, K H [International Atomic Energy Agency, Laboratory Seibersdorf, Vienna (Austria)

    1974-12-01

    In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the {sup 103}Rh(n,n'){sup 103m}Rh and {sup 58}Ni(n,p){sup 58}Co integral cross sections have been accurately measured relatively to the {sup 115}In(n,n'){sup 115m} In cross section in the {sup 235}U thermal fission neutron spectrum and in the MOL-{sigma}{sigma} intermediate-energy standard neutron field. In this last neutron field, the data are related also to the {sup 235}U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific {sup 103}Rh(n,n'){sup 103m}Rh differential-energy cross section among the existing, conflicting data. (author)

  3. Measurements of Integral Cross Section Ratios in Two Dosimetry Benchmark Neutron Fields

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A. [CEN-SCK, Mol (Belgium); Czock, K. H. [International Atomic Energy Agency, Vienna (Austria)

    1974-12-15

    In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the {sup 103}Rh(n,n'){sup 103m}Rh and {sup 58}Ni(n,p){sup 58}Co integral cross sections have been accurately measured relatively to the {sup 115}In(n,n'){sup 115m}In cross section in the {sup 235}U thermal dission neutron spectrum and in the MOL{Sigma}{Sigma} Intermediate-Energy Standard Neutron field. In this last neutron field, the data are related also to the {sup 235}U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific {sup 103}Rh(n,n'){sup 103m}Rh differential-energy cross section among the existing, conflicting data. (author)

  4. Cross section and method uncertainties: the application of sensitivity analysis to study their relationship in radiation transport benchmark problems

    International Nuclear Information System (INIS)

    Weisbi, C.R.; Oblow, E.M.; Ching, J.; White, J.E.; Wright, R.Q.; Drischler, J.

    1975-08-01

    Sensitivity analysis is applied to the study of an air transport benchmark calculation to quantify and distinguish between cross-section and method uncertainties. The boundary detector response was converged with respect to spatial and angular mesh size, P/sub l/ expansion of the scattering kernel, and the number and location of energy grid boundaries. The uncertainty in the detector response due to uncertainties in nuclear data is 17.0 percent (one standard deviation, not including uncertainties in energy and angular distribution) based upon the ENDF/B-IV ''error files'' including correlations in energy and reaction type. Differences of approximately 6 percent can be attributed exclusively to differences in processing multigroup transfer matrices. Formal documentation of the PUFF computer program for the generation of multigroup covariance matrices is presented. (47 figures, 14 tables) (U.S.)

  5. Photoionization cross section of atomic and molecular oxygen

    International Nuclear Information System (INIS)

    Pareek, P.N.

    1983-01-01

    Photoionization cross sections of atomic oxygen and dissociative photoionization cross sections of molecular oxygen were measured from their respective thresholds to 120 angstrom by use of a photoionization mass spectrometer in conjunction with a spark light source. The photoionization cross sections O 2 + parent ion and O + fragment ion from neutral O 2 were obtained by a technique that eliminated the serious problem of identifying the true abundances of O + ions. These ions are generally formed with considerable kinetic energy and, because most mass spectrometers discriminate against energetic ions, true O + abundances are difficult to obtain. In the present work the relative cross sections for producing O + ions are obtained and normalized against the total cross sections in a spectral region where dissociative ionization is not possible. The fragmentation cross sections for O + were then obtained by subtraction of O 2 + cross sections from the known total photoionization cross sections. The results are compared with the previously published measurements. The absolute photoionization cross section of atomic oxygen sigma 8 /sub +/ was measured at 304 A. The actual number density of oxygen atoms within the ionization region was obtained by measuring the fraction of 0 2 molecules dissociated. This sigma/sub +/ at 304 angstrom was used to convert the relative photoinization cross sections, measured as a function of wavelength using a calibrated photodiode, to absolute cross sections. The results are compared with previous measurements and calculated cross sections. angstrom Rydberg series converging to the OII 4 P state was observed

  6. Multilevel parametrization of fissile nuclei resonance cross sections

    International Nuclear Information System (INIS)

    Lukyanov, A.A.; Kolesov, V.V.; Janeva, N.

    1987-01-01

    Because the resonance interference has an important influence on the resonance structure of neutron cross sections energy dependence at lowest energies, multilevel scheme of the cross section parametrization which take into account the resonance interference is used for the description with the same provisions in the regions of the interferential maximum and minimum of the resonance cross sections of the fissile nuclei

  7. Comparison of integral cross section values of several cross section libraries in the SAND-II format

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.

    1978-01-01

    A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

  8. Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method

    International Nuclear Information System (INIS)

    Zheng, S.H.

    1994-01-01

    It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ''benchmarks'' have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section's fluctuation. In this thesis, we propose a new method called ''Probability Table Method'' to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup's results and probability table's results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs

  9. A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

    International Nuclear Information System (INIS)

    Basha, H.S.; Manahan, M.P.

    1992-01-01

    In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

  10. Graphs of the cross sections in the recommended Monte Carlo cross-section library at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Soran, P.D.; Seamon, R.E.

    1980-05-01

    Graphs of all neutron cross sections and photon production cross sections on the Recommended Monte Carlo Cross Section (RMCCS) library have been plotted along with local neutron heating numbers. Values for anti ν, the average number of neutrons per fission, are also given

  11. Graphs of the cross sections in the Alternate Monte Carlo Cross Section library at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Seamon, R.E.; Soran, P.D.

    1980-06-01

    Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes

  12. Accurate measurements of neutron activation cross sections

    International Nuclear Information System (INIS)

    Semkova, V.

    1999-01-01

    The applications of some recent achievements of neutron activation method on high intensity neutron sources are considered from the view point of associated errors of cross sections data for neutron induced reaction. The important corrections in -y-spectrometry insuring precise determination of the induced radioactivity, methods for accurate determination of the energy and flux density of neutrons, produced by different sources, and investigations of deuterium beam composition are considered as factors determining the precision of the experimental data. The influence of the ion beam composition on the mean energy of neutrons has been investigated by measurement of the energy of neutrons induced by different magnetically analysed deuterium ion groups. Zr/Nb method for experimental determination of the neutron energy in the 13-15 MeV energy range allows to measure energy of neutrons from D-T reaction with uncertainty of 50 keV. Flux density spectra from D(d,n) E d = 9.53 MeV and Be(d,n) E d = 9.72 MeV are measured by PHRS and foil activation method. Future applications of the activation method on NG-12 are discussed. (author)

  13. [Fast neutron cross section measurements]: Progress report

    International Nuclear Information System (INIS)

    1988-01-01

    As projected in our previous proposal, the past year on the cross section project at the University of Michigan has been one primarily of construction and assembly of our 14 MeV pulsed Neutron Facility. All the components of the system have now been either purchased or fabricated in our shop facilities and have been assembled in their final configuration. We are now in the process of testing the rf components that have been designed to deliver voltage to both the pulser and buncher stages. We expect that the system will be operational by the end of the current contract year. We have also accomplished the design and construction of several other major pieces of equipment that are needed to begin fast neutron time-of-flight measurements. These include the primary proton recoil detector, and a californium fission chamber needed in the efficiency calibration of the primary detector. We have also added considerable concrete shielding designed to lower the neutron background in the experimental area. 10 figs., 5 tabs

  14. Neutron cross section measurements at ORELA

    International Nuclear Information System (INIS)

    Dabbs, J.W.T.

    1979-01-01

    ORELA (Oak Ridge Electron Linear Accelerator) has been for the last decade the most powerful and useful pulsed neutron time-of-flight facility in the world, particularly in the broad midrange of neutron energies (10 eV to 1 MeV). This position will be enhanced with the addition of a pulse narrowing prebuncher, recently installed and now under test. Neutron capture, fission, scattering, and total cross sections are measured by members of the Physics and Engineering Physics Divisions of ORNL, and by numerous guests and visitors. Several fundamental and applied measurements are described, with some emphasis on instrumentation used. The facility comprises the accelerator and its target(s), 10 evacuated neutron flight paths having 18 measurement stations at flight path distances 8.9 to 200 meters, and a complex 4-computer data acquisition system capable of handling some 17,000 32-bit events/s from a total of 12 data input ports. The system provides a total of 2.08 x 10 6 words of data storage on 3 fast disk units. In addition, a dedicated PDP-10 timesharing system with a 250-megabyte disk system and 4 PDP-15 graphic display satellites permits on-site data reduction and analysis. More than 10 man-years of application software development supports the system, which is used directly by individual experiments. 12 figures, 1 table

  15. Resonance capture cross section of 207Pb

    CERN Document Server

    Domingo-Pardo, C.; Aerts, G.; Alvarez-Pol, H.; Alvarez-Velarde, F.; Andriamonje, S.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; Badurek, G.; Baumann, P.; Becvar, F.; Berthoumieux, E.; Bisterzo, S.; Calvino, F.; Cano-Ott, D.; Capote, R.; Carrapico, C.; Cennini, P.; Chepel, V.; Chiaveri, E.; Colonna, N.; Cortes, G.; Couture, A.; Cox, J.; Dahlfors, M.; David, S.; Dillman, I.; Dolfini, R.; Dridi, W.; Duran, I.; Eleftheriadis, C.; Embid-Segura, M.; Ferrant, L.; Ferrari, A.; Ferreira-Marques, R.; Fitzpatrick, L.; Frais-Koelbl, H.; Fujii, K.; Furman, W.; Gallino, R.; Goncalves, I.; Gonzalez-Romero, E.; Goverdovski, A.; Gramegna, F.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Haas, B.; Haight, R.; Heil, M.; Herrera-Martinez, A.; Igashira, M.; Isaev, S.; Jericha, E.; Kadi, Y.; Kappeler, F.; Karamanis, D.; Karadimos, D.; Kerveno, M.; Ketlerov, V.; Koehler, P.; Konovalov, V.; Kossionides, E.; Krticka, M.; Lamboudis, C.; Leeb, H.; Lindote, A.; Lopes, I.; Lozano, M.; Lukic, S.; Marganiec, J.; Marrone, S.; Mastinu, P.; Mengoni, A.; Milazzo, P.M.; Moreau, C.; Mosconi, M.; Neves, F.; Oberhummer, H.; Oshima, M.; O'Brien, S.; Pancin, J.; Papachristodoulou, C.; Papadopoulos, C.; Paradela, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Perrot, L.; Plag, R.; Plompen, A.; Plukis, A.; Poch, A.; Pretel, C.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rosetti, M.; Rubbia, C.; Rudolf, G.; Rullhusen, P.; Salgado, J.; Sarchiapone, L.; Savvidis, I.; Stephan, C.; Tagliente, G.; Tain, J.L.; Tassan-Got, L.; Tavora, L.; Terlizzi, R.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vincente, M.C.; Vlachoudis, V.; Vlastou, R.; Voss, F.; Walter, S.; Wendler, H.; Wiescher, M.; Wisshak, K.

    2006-01-01

    The radiative neutron capture cross section of 207Pb has been measured at the CERN neutron time of flight installation n_TOF using the pulse height weighting technique in the resolved energy region. The measurement has been performed with an optimized setup of two C6D6 scintillation detectors, which allowed us to reduce scattered neutron backgrounds down to a negligible level. Resonance parameters and radiative kernels have been determined for 16 resonances by means of an R-matrix analysis in the neutron energy range from 3 keV to 320 keV. Good agreement with previous measurements was found at low neutron energies, whereas substantial discrepancies appear beyond 45 keV. With the present results, we obtain an s-process contribution of 77(8)% to the solar abundance of 207Pb. This corresponds to an r-process component of 23(8)%, which is important for deriving the U/Th ages of metal poor halo stars.

  16. Josephson cross-sectional model experiment

    International Nuclear Information System (INIS)

    Ketchen, M.B.; Herrell, D.J.; Anderson, C.J.

    1985-01-01

    This paper describes the electrical design and evaluation of the Josephson cross-sectional model (CSM) experiment. The experiment served as a test vehicle to verify the operation at liquid-helium temperatures of Josephson circuits integrated in a package environment suitable for high-performance digital applications. The CSM consisted of four circuit chips assembled on two cards in a three-dimensional card-on-board package. The chips (package) were fabricated in a 2.5-μm (5-μm) minimum linewidth Pb-alloy technology. A hierarchy of solder and pluggable connectors was used to attach the parts together and to provide electrical interconnections between parts. A data path which simulated a jump control sequence and a cache access in each machine cycle was successfully operated with cycle times down to 3.7 ns. The CSM incorporated the key components of the logic, power, and package of a prototype Josephson signal processor and demonstrated the feasibility of making such a processor with a sub-4-ns cycle time

  17. Female medical leadership: cross sectional study.

    Science.gov (United States)

    Kvaerner, K J; Aasland, O G; Botten, G S

    1999-01-09

    To assess the relation between male and female medical leadership. Cross sectional study on predictive factors for female medical leadership with data on sex, age, specialty, and occupational status of Norwegian physicians. Oslo, Norway. 13 844 non-retired Norwegian physicians. Medical leaders, defined as physicians holding a leading position in hospital medicine, public health, academic medicine, or private health care. 14.6% (95% confidence interval 14.0% to 15.4%) of the men were leaders compared with 5.1% (4.4% to 5.9%) of the women. Adjusted for age men had a higher estimated probability of leadership in all categories of age and job, the highest being in academic medicine with 0.57 (0.42 to 0.72) for men aged over 54 years compared with 0.39 (0.21 to 0.63) for women in the same category. Among female hospital physicians there was a positive relation between the proportion of women in their specialty and the probability of leadership. Women do not reach senior positions as easily as men. Medical specialties with high proportions of women have more female leaders.

  18. New resonance cross section calculational algorithms

    International Nuclear Information System (INIS)

    Mathews, D.R.

    1978-01-01

    Improved resonance cross section calculational algorithms were developed and tested for inclusion in a fast reactor version of the MICROX code. The resonance energy portion of the MICROX code solves the neutron slowing-down equations for a two-region lattice cell on a very detailed energy grid (about 14,500 energies). In the MICROX algorithms, the exact P 0 elastic scattering kernels are replaced by synthetic (approximate) elastic scattering kernels which permit the use of an efficient and numerically stable recursion relation solution of the slowing-down equation. In the work described here, the MICROX algorithms were modified as follows: an additional delta function term was included in the P 0 synthetic scattering kernel. The additional delta function term allows one more moments of the exact elastic scattering kernel to be preserved without much extra computational effort. With the improved synthetic scattering kernel, the flux returns more closely to the exact flux below a resonance than with the original MICROX kernel. The slowing-down calculation was extended to a true B 1 hyperfine energy grid calculatn in each region by using P 1 synthetic scattering kernels and tranport-corrected P 0 collision probabilities to couple the two regions. 1 figure, 6 tables

  19. Sensitivity study and functionalization of cross section to fuel and moderator temperature

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Song, Jae Seung; Cho, Young Chul

    1995-11-01

    A reactor core neutronics code MASTER is under development as a part of Korean Core Design System ADONIS. MASTER solves two-group three-dimensional; neutron diffusion equation which requires fuel assembly-wise group constants, to calculate the neutron flux distribution in the core. The group constants are obtained from the fuel assembly multi-group neutron transport calculation, and inputted as functions of the core operating condition. The functionalization of the group constant requires sensitivity analysis to various core operating conditions. In this report, the sensitivity of group constant to fuel and moderator temperature were analyzed. Lumped higher order macroscopic cross section derivative method was developed to reduce the computer memory and the number of floating point operations to treat group constants in MASTER. 1 fig., 6 tabs., 2 refs. (Author) .new

  20. Preparation of next generation set of group cross sections. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, based on the examination result about the evaluation energy range of heavy element unresolved resonance cross sections, the upper energy limit of the energy range, where ultra-fine group cross sections are produced, was raised to 50 keV, and an improvement of the group cross section processing system was promoted. At the same time, reflecting the result of studies carried out till now, a function producing delayed neutron data was added to the general-purpose group cross section processing system , thus the preparation of general purpose group cross section processing system has been completed. On the other hand, the energy structure, data constitution and data contents of next generation group cross section set were determined, and the specification of a 151 groups next generation group cross section set was defined. Based on the above specification, a concrete library format of the next generation cross section set has been determined. After having carried out the above-described work, using the general-purpose group cross section processing system , which was complete in this study, with use of the JENDL-3. 2 evaluated nuclear data, the 151 groups next generation group cross section of 92 nuclides and the ultra fine group resonance cross section library for 29 nuclides have been prepared. Utilizing the 151 groups next generation group cross section set and the ultra-fine group resonance cross-section library, a bench mark test calculation of fast reactors has been performed by using an advanced lattice calculation code. It was confirmed, by comparing the calculation result with a calculation result of continuous energy Monte Carlo code, that the 151 groups next generation cross section set has sufficient accuracy. (author)