S. Ganesan

GEANT4 simulation of the neutron background of the C6D6 set-up for capture studies at n_TOF

The neutron sensitivity of the C6D6 detector setup used at n TOF for capture measurements has been studied by means of detailed GEANT4 simulations. A realistic software replica of the entire n TOF experimental hall, including the neutron beam line, sample, detector supports and the walls of the experimental area has been implemented in the simulations. The simulations have been analyzed in the same manner as experimental data, in particular by applying the Pulse Height Weighting Technique. The simulations have been validated against a measurement of the neutron background performed with a nat C sample, showing an excellent agreement above 1 keV. At lower energies, an additional component in the measured nat C yield has been discovered, which prevents the use of nat C data for neutron background estimates at neutron energies below a few hundred eV. The origin and time structure of the neutron background have been derived from the simulations. Examples of the neutron background for two di erent samples are demonstrating the important role of accurate simulations of the neutron background in capture cross section measurements.

New investigations of the criticality property of pure 232U

Abstract This paper presents results of new calculations of criticality of 232 U performed using the basic evaluated nuclear data files JENDL-3.2 (Japan) and ENDF/B-VI.5 (USA) using the NJOY-MCNP code system. Comparisons of these two basic nuclear data files have been done using PREPRO2000 code system and are presented. The critical mass of 232 U calculated using ENDF/B-VI.5 and JENDL-3.2 respectively are 3.73 and 13.6 kg. The paper presents a mention of formation routes of 232 U. A few remarks on the role of 232 U in providing resistance to proliferation of fissile material with respect to utilizing thorium in thermal, fast, fusion and accelerator driven systems are made.

Fission Product Yield in the Neutron-Induced Fission of 232Th with Average Energies of 5.42, 7.75, and 10.09 MeV

Abstract The yields of various fission products in the neutron-induced fission of 232Th have been determined a using recoil catcher and off-line gamma-ray spectrometric technique with flux-averaged energies of 5.42, 7.75, and 10.09 MeV. The neutrons were generated using the 7Li(p,n) reaction at the BARC-TIFR [Bhabha Atomic Research Centre–Tata Institute of Fundamental Research] Pelletron facility, Mumbai, India. The fission product–yield data in the 10.09-MeV neutron-induced fission of 232Th are determined for the first time. The yields of the different fission products in the neutron-induced fission of 232Th with flux-averaged energies of 5.42 and 7.75 MeV from the present work have been compared with similar data of comparable neutron energy from the literature and are found to be in good agreement. The effect of nuclear structure on fission product yields as a function of neutron energy has been examined.

Measurement of Neutron-Induced Reaction Cross Sections in Zirconium Isotopes at Thermal, 2.45 MeV and 9.85 MeV Energies

Abstract The 94Zr(n,γ)95Zr and 90Zr(n,p)90Ym reaction cross sections were measured at neutron energies En of 2.45 MeV and 9.85 ± 0.38 MeV (average) using an activation and off-line gamma-ray spectrometric technique. In addition to these, the thermal neutron capture cross sections of 94Zr(n,γ)95Zr and 96Zr(n,γ)97Zr were also measured using the same technique. The experimentally measured neutron cross-section data were compared with the latest available evaluated nuclear data libraries from ENDF/B-VII, JENDL 4.0, and TENDL 2010.

Self-shielding and energy dependence of dilution cross-section in the resolved resonance region

Abstract In the calculation of flux weighted multigroup neutron cross-sections of a nuclide present in a homogeneous mixture of several nuclides, by the well known Bondarenko approach, the contributions from the other nuclides to the flux weighting function, are contained in a parameter called ‘dilution’. Under the narrow resonance approximation and with assumed non-overlap of resonances, the energy fluctuations of the dilution are usually ignored within an energy group. The ‘self-shielding factor’ (SSF), which is the ratio of the group cross section for a given dilution to that at infinite dilution, is a smooth function of dilution. A conventional multigroup cross section set, apart from the infinite dilution cross-sections, gives SSF over a dilution grid, from which the effective SSF for any dilution could be obtained by interpolation. The effective SSF multiplied by the infinite dilution cross-section then gives the effective cross-section to be used in the neutronic analyis. Though this conventional procedure works well, the accuracy of this procedure depends on the interpolation scheme used and on the fineness of the dilution grid. The effect of fineness of the dilution grid and the effect of energy dependence of the dilution on the effective SSF are observed, for a specific case, and the results presented in this paper. The JENDL-2 basic nuclear data was used and the benchmark fast critical assembly ZPR-6-7 analysed for a temperature of 300K. The study was restricted to the resolved resonance regions of the nuclides involved. The results appear significant with respect to the target accuracies demanded for the SSF.

Generation of handbook of multi-group cross sections of WIMS-D libraries by using the XnWlup2.0 software

Abstract A project to prepare an exhaustive handbook of WIMS-D cross section libraries for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully designed. To meet the objectives of this project, a computer software package with graphical user interface for MS Windows has been developed at BARC, India. This article summarizes the salient features of this new software and presents significant improvements and extensions in relation to its first version [Ann Nucl Energ 29 (2002) 1735].

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

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.

Computational Schemes for Online Flux Mapping System in a Large-Sized Pressurized Heavy Water Reactor

Abstract Large-sized pressurized heavy water reactors (PHWRs) are neutronically loosely coupled and hence are prone to significant changes in flux shape during operation. As a result, they need a sophisticated regulation procedure based on an online flux mapping system (OFMS). During the reactor operation, neutron flux is continuously measured at certain predetermined in-core locations. The purpose of OFMS is to compute a detailed flux map at all points in the reactor, after every 2 min, by making use of the measured fluxes. The knowledge of detailed flux distribution is then used for an appropriate regulating action. The choice of computational method used by OFMS is of crucial importance because the method is expected to be both efficient and accurate and should work for a range of reactor configurations occurring during the operation. In this paper, three different methods, namely, flux synthesis, internal boundary condition, and combined least squares (CLSQ), are analyzed for their prospective use in the forthcoming 700-MW(electric) Indian PHWR. The CLSQ method is found to be most accurate, although it needs significant computation. A hybrid method that combines certain features of other methods is also studied and seems to give good accuracy with moderate computational effort.

Measurement of formation cross-section of 99Mo from the 98Mo(n,γ) and 100Mo(n,2n) reactions

The formation cross-section of medical isotope 99Mo from the 98Mo(n,γ) reaction at the neutron energy of 0.025eV and from the 100Mo(n,2n) reaction at the neutron energies of 11.9 and 15.75MeV have been determined by using activation and off-line γ-ray spectrometric technique. The thermal neutron energy of 0.025eV was used from the reactor critical facility at BARC, Mumbai, whereas the average neutron energies of 11.9 and 15.75MeV were generated using 7Li(p,n) reaction in the Pelletron facility at TIFR, Mumbai. The experimentally determined cross-sections were compared with the evaluated nuclear data libraries of ENDF/B-VII.1, CENDL-3.1, JENDL-4.0 and JEFF-3.2 and are found to be in close agreement. The 100Mo(n,2n)99Mo reaction cross-sections were also calculated theoretically by using TALYS-1.8 and EMPIRE-3.2 computer codes and compared with the experimental data.

Fission Neutron Spectrum Sensitivity Study for the Case of Advanced Heavy Water Reactor

Abstract In the WIMSD-IAEA multigroup nuclear data library, the isotopes and weights adopted for WLUP libraries to calculate the average fission spectra for 235U, 238U, and 239Pu are in the ratio of 54%, 8%, and 38%, respectively. The average fission neutron spectrum in the existing multigroup WIMSD-IAEA library applicable for the U-Pu cycle is not rigorously applicable for systems that are based on the thorium fuel cycle because of two aspects. First, the weightage of the fission neutron spectrum of 232Th and 233U nuclides, which are important isotopes in the thorium fuel cycle, are not considered in obtaining the average multigroup fission spectrum in the conventional WIMSD-IAEA library. Second, the 232Th/233U system spectrum is required for condensation of the fission spectrum as done in generating other multigroup cross sections and parameters for the thorium fuel cycle. In this work, we have processed the fission neutron spectrum data from the basic evaluated nuclear data file (ENDF/B-VI.8) for each important isotope in the thorium fuel cycle using the Th/233U spectrum and using a FORTRAN program developed and validated by us for this purpose. The final average fission spectrum to be fed into the WIMSD-IAEA library is prepared by mixing the isotopic multigroup fission spectrum of individual isotopes 233U, 239Pu, and 241Pu with appropriate weights corresponding to their respective power fractions in the advanced heavy water reactor (AHWR) lattice. Using the WIMSD library with modified effective fission spectra, the lattice k-infinity calculations of AHWR are performed as a function of burnup. The difference in the infinite multiplication factor, which is expressed in terms of reactivity in mk, ranges from 0.48 to 0.94 mk as burnup in the AHWR proceeds from 0 to 55 GWd/tonne.