Robert C. Little

Benchmark Calculations for the Doppler Coefficient of Reactivity

This paper reports on the Doppler coefficient of reactivity that is a crucial parameter in the evaluation of several transients in light water reactors (LWRs). It is relatively small in magnitude and cannot be measured directly in operating reactors. Doppler coefficients are presented for slightly idealized pressurized water reactor pin cells. These coefficients were calculated with the MCNP-3A continuous-energy Monte Carlo code using data taken directly from the ENDF/B-V nuclear data library. This combination represents the most rigorous analytical tool and the best nuclear data available. Consequently, these results comprise a set of numerical benchmarks that may be used to evaluate the accuracy of LWR lattice physics codes in predicting Doppler behavior at operating conditions. An example of one such evaluation, using the CELL-2 code, is included.

Evaluation and propagation of the 239Pu fission cross-section uncertainties using a monte carlo technique

Abstract We present an approach to uncertainty quantification for nuclear applications that combines the covariance evaluation of differential cross-section data and the error propagation from matching a criticality experiment using a neutron-transport calculation. We have studied the reduction in uncertainty of 239Pu fission cross sections by using a one-dimensional neutron-transport calculation with the PARTISN code. The evaluation of 239Pu differential cross-section data is combined with a criticality measurement (Jezebel) using a Bayesian method. To quantify the uncertainty in such calculations, we generate a set of random samples of the cross sections, which represents the covariance matrix, and estimate the distribution of calculated quantities, such as criticality. We show that inclusion of the Jezebel data reduces uncertainties in estimating neutron multiplicity.

Two-dimensional discrete ordinates photon transport calculations for brachytherapy dosimetry applications

Abstract The DANTSYS discrete ordinates computer code system is applied to quantitative estimation of water kerma rate distributions in the vicinity of discrete photon sources with energies in the 20- to 800-keV range in two-dimensional cylindrical r-z geometry. Unencapsulated sources immersed in cylindrical water phantoms of 40-cm diameter and 40-cm height are modeled in either homogeneous phantoms or shielded by Ti, Fe, and Pb filters with thicknesses of 1 and 2 mean free paths. The obtained dose results are compared with corresponding photon Monte Carlo simulations. A 210-group photon cross-section library for applications in this energy range is developed and applied, together with a general-purpose 42-group library developed at Los Alamos National Laboratory, for DANTSYS calculations. The accuracy of DANTSYS with the 42-group library relative to Monte Carlo exhibits large pointwise fluctuations from –42 to +84%. The major cause for the observed discrepancies is determined to be the inadequacy of the weighting function used for the 42-group library derivation. DANTSYS simulations with a finer 210-group library show excellent accuracy on and off the source transverse plane relative to Monte Carlo kerma calculations, varying from –4.9 to 3.7%. The P3 Legendre polynomial expansion of the angular scattering function is shown to be sufficient for accurate calculations. The results demonstrate that DANTSYS is capable of calculating photon doses in very good agreement with Monte Carlo and that the multigroup cross-section library and efficient techniques for mitigation of ray effects are critical for accurate discrete ordinates implementation.

Uncertainty Quantification of Prompt Fission Neutron Spectrum for n(0.5 MeV) + 239Pu

Abstract Uncertainties associated with the prompt fission neutron spectrum (PFNS) of n(0.5 MeV) + 239Pu evaluated for the ENDF/B-VII.0 library are estimated using known experimental information and model parameter uncertainties in the framework of the Madland-Nix model. The model parameters used for the ENDF/B-VII.0 evaluation are also used in the present work. A covariance matrix is obtained, and its eigenvalues are estimated. Sampled spectra are then used in PARTISN transport simulations to infer the impact of PFNS uncertainties on the calculation of the multiplication factor keff in the Jezebel critical assembly. The present evaluated PFNS uncertainties lead to ˜0.24% uncertainty in the Jezebel keff. Finally, multigroup covariance matrices are produced in 33- and 590-group structures.

Nuclear data for accelerator-driven systems

Abstract We review recent evaluations of neutron and proton reaction cross sections up to 150 MeV in the LA150 Library, for use in computer code simulations of accelerator-driven systems. An overview is provided of the nuclear theory together with measured cross section data used in the evaluations. The possible use of bismuth activation foils for high-energy neutron spectrometry is also discussed. We describe recent developments to the MCNPX radiation transport code, which merges MCNP and LAHET in one code and uses the LA150 evaluated data. A number of benchmark comparisons against integral experiments are described, for thick-target neutron production, neutron transmission through macroscopic slabs, and neutron kerma coefficients. The benchmarks help validate the transport code and the evaluated data for use in ADS simulations of neutron production in a spallation target (n/p), radiation shielding, heating, and damage. A brief summary is also given of future data needs for subcritical transmuters and spallation targets, in accelerator transmutation of waste technologies.

ENDF70: A Continuous-Energy MCNP Neutron Data Library Based on ENDF/B-VII.0

Abstract Following the release of ENDF/B-VII.0 evaluations, an ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced at Los Alamos National Laboratory. This new library contains data for 387 isotopes and three elements at five temperatures: 293.6, 600, 900, 1200, and 2500 K. It can be obtained as part of the MCNP5 1.50 release. The new library was created using ENDF/B-VII.0 neutron evaluations and primarily version 248 of NJOY99. A processing script was created that set up the input files for NJOY and employed checking codes to test the content of the processed data. A sample MCNP run was performed for each isotope and temperature, and cross sections for each isotope were plotted to make sure there were no major problems. The processed ACE libraries did not always pass all quality assurance tests. For example, energy-balance problems were identified for several evaluations having negative heating numbers or inconsistencies between total and partial heating. Similarly, some problems were found with unresolved resonance probability tables, resulting in probability tables being excluded from the final library for several materials. Certain evaluations were modified and reprocessed as a result of the quality assurance tests, and some data points in the final ACE files were changed because they were too small or had other problems. The new ENDF70 library provides MCNP users with the latest ENDF/B data available. This collection of data includes a larger range of isotopes and temperatures than previously released, which will be beneficial in numerous applications. The upgrades included as part of ENDF/B-VII.0 and, hence, ENDF70 should improve calculations.