摘要:

AbstractAn approach is summarized for developing a full analytical method for the generation of laboratory (lab) coordinate system multigroup transfer cross sections of elastic and discrete-level inelastic scatterings of neutrons, where the angular distribution data of the scattered neutrons are given as coefficients of truncated Legendre polynomial expansions in the center-of-mass (c.m.) coordinate system. In the“kernel form”of the multigroup approximation, fluxes, cross sections, and angular data are left outside the integration signs of the transfer cross-section expression. Then, the integrand is a four-index kernel-the source and sink energy groups and the Legendre polynomials in the c.m. and lab systems each contributing one index—integrated over the source and sink groups. In the method introduced, the double integration on the neutron pre- and postscattering energies, in these two groups, is carried out analytically. This is done by decomposing the Legendre polynomials in the integrand into their representations as polynomials of their arguments—the cosines of the angle of scattering in the lab and the c.m. systems. It is assumed that the flux is presented as a simple function of the neutron prescattering energy. The first integration takes place between the cosines of scattering in the c.m. system related by the kinematics of scattering to the energy boundaries of the sink group, while the second integration takes place between the energy boundaries of the source group. Transfer cross sections of discrete-level inelastic scatterings are evaluated quite similarly to those of elastic scatterings. The only difference is in the integration over the prescattering energies, which stems from the replacement of the mass number of the scattering nucleus, appearing in the expression for the cosine of scattering in the lab system, by an energy-dependent“effective mass number”in the case of inelastic scattering. The method of evaluation described seems to be fast and accurate and is valid theoretically up to any order of Legendre expansions desired. Some evaluations made with this method, for both elastic and discrete-level inelastic scatterings, are compared with numerical computations carried out by other methods: the NJOY cross-section processing code and the method outlined by Hong and Shultis. The flux used as a weighting function is either energy independent or of a 1/E’form (where E’is the prescattering energy of the neutron). The results are in good agreement.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15482

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractContributons are the special particles distributed among a general population that generate the response observed on a specified detector. Contributon slowing-down theory describes the transfer of the response through space and energy as it is carried by contributons from the source to the detector. The response flow through space-energy and space-lethargy obeys the contributon slowing-down equation, which expresses conservation of contributons. A four-dimensional vector field is introduced to identify space and energy channels followed by the contributons, and is used to define response flow lines through space-lethargy. Numerical expressions are presented to compute the response current and slowing-down density that define the components of the response flow field. It is shown how these variables can be used to perform energy channel theory analysis of a particle transport problem. The method is applied to two realistic problems. The first determines contributon transport channels followed through space-energy by fission neutrons produced in a pressurized water reactor as they travel from the core to the reactor cavity region, where they activate surveillance dosimeters. The second examines the response transfer from a nuclear weapon burst as it is carried by contributons through space-lethargy channels in air to detectors located at some distance from ground-zero.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15483

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractAs a basic study of photonuclear transmutation of actinides in high-level radioactive wastes using electron-produced bremsstrahlung, the absolute yields of cumulative mass distributions and the transmutation rates of235U,238U,237Np, and239Pu by photofission reactions induced by 20-, 30-, and 60-MeV bremsstrahlung were measured. The results of mass yield distributions and transmutation yields agree well with other experimental results and those calculated using photofission cross sections, respectively. The transmutation efficiency per electron increases about one order of magnitude with electron energy from 20 to 60 MeV.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15484

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractDirect energy transfer by fission fragments near the wall of a cavity containing fissioning gas is studied in plane and cylindrical geometries. Analytical formulas are derived for the fission fragment energy flux. Heat transfer equations are solved for optically thick fissioning gases by taking into account the fission fragment energy transport effect. The results are applied to a heat transfer analysis of the fuel assemblies of a heterogeneous gas core reactor. The energy transfer mechanism in the fissioning gas is essentially nonlinear. Thus, the cooling effect due to direct fission fragment energy loss to the container walls does not become significant until the stopping range considerably exceeds the characteristic dimensions of the container. For example, when the ratio of the stopping range to the container dimensionλ/δis equal to 3, 45% of the energy flux at the container walls is due to the fission fragments; yet the maximum fuel temperature decreases by only l0%. If the ratioλ/δis∼100, fission fragments account for 95% of the energy flux to the walls, and the gas temperature decreases by 50%.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15485

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractAn intense semimonoenergetic neutron field was made using a simple beryllium target system bombarded by protons of nine different energies between 20 and 40 MeV. Natural sodium, aluminum, vanadium, chromium, manganese, copper, zinc, and gold samples were irradiated at this field, and gamma rays from the samples were observed by a germanium detector. The production rates of 17 radionuclides were obtained for the nine different neutron fields, and the excitation functions of these 17 reaction channels of23Na(n,2n)22Na,27Al(n,α)24Na,51V(n,α)48Sc,51V(n,p)51Ti, 50Cr(n,3n)48Cr,50Cr(n,2n)49Cr,55Mn(n,4n)51Ti,55Mn(n,4n)52Mn,55Mn(n,2n)54Mn,63Cu(n,3n) Cu,63Cu(n,2n)62Cu,65Cu(n,p)65Ni,64Zn(n,t)62Cu,64Zn(n,3n)62Zn,64Zn(n,2n)63Zn,197Au(n,4n)194Au, and197Au(n,2n)196Au were obtained for neutron energies up to 40 MeV by using the SAND-II and the NEUPAC unfolding codes and also least-squares fitting. The initial guess value for these methods was obtained primarily from calculations of the ALICE/LIVERMORE82 code.

ISSN:0029-5639    

DOI:10.13182/NSE111-391

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractAn evaluation of the238U neutron cross sections in the unresolved resonance region that was adopted for the evaluated nuclear data files JEF-2 (up to 200 keV) and ENDF/B-VI (up to 149 keV) has been checked against recent capture cross-section measurements and against thick-sample transmission data and capture self-indication ratios. Effects of the unresolved resonance structure on self-shielding and multiple scattering were treated by Monte Carlo techniques based on resonance statistics and average resonance parameters. It was found that the average cross sections and average resonance parameters given in the new evaluation permit very satisfactory reproduction of all the test data. Indications are that the average total and capture cross sections including self-shielding are now known below 200 keV with accuracies close to those requested in nuclear technology.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15487

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractThe fission cross sections of235U and239Pu are measured with very high neutron energy resolution (0.17 ns/m) in the energy region from 100 to 2000 eV for235U and to 20000 eV for239Pu. The purpose of this measurement is to provide fission cross sections with energy resolution comparable with that available from transmission measurements for the purpose of deriving multilevel resolved resonance parameters. Fission ion chambers are used to detect fission fragments, and a10B ionization chamber is used to measure the relative neutron flux at the 86-m flight path of the Oak Ridge Electron Linear Accelerator. The measured fission cross sections are the highest resolution measurements of good accuracy reported in the neutron energy range above 400 eV.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15488

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

摘要:

AbstractA constrained least-squares method is developed for deducing six-group delayed neutron energy spectra from composite spectra measured at six or more delay time intervals following fission. The constraining condition is chosen to yield stable solutions that also provide good fits to the measured spectra. The method is applied to previously measured composite spectra of235U to obtain six-group delayed neutron energy spectra. The solutions are unique for a large range of constraint spectra. The dependence of the solutions on the choice of six-group parameters (βj,λj) is also examined.

ISSN:0029-5639    

DOI:10.13182/NSE92-A15489

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor

ISSN:0029-5639    

DOI:10.13182/NSE111-433

出版商:Taylor&Francis    

年代:1992

数据来源: Taylor