Morgan C. White

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.

Photonuclear Physics in Radiation Transport - I: Cross Sections and Spectra

Abstract This paper describes model calculations and nuclear data evaluations of photonuclear reactions on isotopes of C, O, Al, Si, Ca, Fe, Cu, Ta, W, and Pb for incident photon energies up to 150 MeV. The calculations, using the GNASH code, include giant-dipole resonance and quasi-deuteron models for photoabsorption. The emission of secondary particles and gamma rays is computed using preequilibrium theory, together with an open-ended sequence of compound nucleus decays using the Hauser-Feshbach theory. The accuracy of the calculated and evaluated cross sections is assessed through extensive comparisons with measured cross sections, average neutron multiplicities, and energy-dependent emission spectra. The evaluated nuclear data files (ENDF) facilitate radiation transport studies of the importance of photonuclear reactions in a number of technologies including photoneutrons produced in electron/photon accelerators, shielding studies, and nondestructive detection of nuclear materials. A companion paper describes developments in the MCNP and MCNPX codes to utilize these data in transport simulations.

Verification of the pulse height tally in MCNP 5

Abstract Pulse height tallies are commonly used in Monte Carlo codes to predict detailed measured photon spectra for spectrometry purposes. The pulse height tally is unique among the various tallies in MCNP. Unlike flux or current tallies, which are calculated as soon as the particle exits or collides in the cell, the entire set of tracks for a history must be completed before the pulse height tally can be made. The objective of this work was to verify the pulse height tally and prepare to verify the new MCNP 5 variance reduction features with the pulse height tally. In this paper, we give details to the analytic solution of the pulse height distribution using a modification to Shuttleworth’s fictitious elements, report MCNP 5 results for the pulse height tally, energy deposited and current tallies for the problem.

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.

Two detector arrays for fast neutrons at LANSCE

The neutron spectrum from neutron-induced fission needs to be known in designing new fast reactors, predicting criticality for safety analyses, and developing techniques for global security application. The experimental data base of fission neutron spectra is very incomplete and most present evaluated libraries are based on the approach of the Los Alamos Model. To validate these models and to provide improved data for applications, a program is underway to measure the fission neutron spectrum for a wide range of incident neutron energies using the spallation source of fast neutrons at the Weapons Neutron Research (WNR) facility at the Los Alamos Neutron Science Center (LANSCE). In a double time-of-flight experiment, fission neutrons are detected by arrays of neutron detectors to increase the solid angle and also to investigate possible angular dependence of the fission neutrons. The challenge is to measure the spectrum from low energies, down to 100 keV or so, to energies over 10 MeV, where the evaporation-like spectrum decreases by 3 orders of magnitude from its peak around 1 MeV. For these measurements, we are developing two arrays of neutron detectors, one based on liquid organic scintillators and the other on 6Li-glass detectors. The range of fission neutrons detected by organic liquid scintillators extends from about 600 keV to well over 10 MeV, with the lower limit being defined by the limit of pulse-shape discrimination. The 6Li-glass detectors have a range from very low energies to about 1 MeV, where their efficiency then becomes small. Various considerations and tests are in progress to understand important contributing factors in designing these two arrays and they include selection and characterization of photomultiplier tubes (PM), the performance of relatively thin (1.8 cm) 6Li-glass scintillators on 12.5 cm diameter PM tubes, use of 17.5 cm diameter liquid scintillators with 12.5 cm PM tubes, measurements of detector efficiencies with tagged neutrons from the WNR/LANSCE neutron beam, and efficiency calibration with 252Cf spontaneous fission neutrons. Design considerations and test results are presented.

Development of Neutron Detector Arrays for Neutron-Induced Reaction Measurements

The outgoing neutron energy spectra from neutron-induced fission of various actinides are important for basic understanding of the fission process near the scission point as well as playing a large role in neutron transport codes, which are heavily relied upon in the design of advanced nuclear reactors and simulations of critical assemblies. The reliability of the results of neutron transport models is a strong function of the quality of the nuclear data used as input. Currently, the worlds experimental database of fission neutron spectra is severely incomplete (especially for higher incident neutron energies) with large uncertainties in key portions of the outgoing energy spectra. Many transport codes use evaluated data libraries, which are based on the approach of the Los Alamos model. Other theoretical models have been developed, but the available data cannot distinguish the results of different models (as is the case for 239Pu). Better measurements are needed for all incident and outgoing neutron energies, but most urgently in the low-energy (below 1 MeV) and high-energy (above 6 MeV) portions of the outgoing spectra where theoretical model results differ greatly. We present the design considerations (and some characterization results) of the two Chi-Nu neutron detector arrays: one array of 6Li-glass detectors and one array of liquid-scintillator detectors. These detector arrays are being constructed to meet the challenge of measuring the prompt fission neutron spectra (for a few common actinides) to a higher accuracy and precision than achieved previously and over a larger incident energy range than has been covered by previous experimenters. We see a significant reduction in neutron-scattering backgrounds with our new array designs.

Meta-Analysis Options for Inconsistent Nuclear Measurements

Abstract Meta-analysis aims to combine results from multiple experiments. For example, a neutron reaction rate or cross section is typically measured in multiple experiments, and a single estimate and its uncertainty are provided for users of the estimated reaction rate. It is often difficult to combine estimates from multiple laboratories because there can be important differences in experimental protocols among laboratories and because laboratories do not always provide all the information needed to assess the estimate’s uncertainty, particularly if total uncertainty (random and systematic) is required. The paper illustrates that explicit measurement error models are essential for understanding measurement processes and for guiding how to combine multiple measurements, whether the measurements are consistent or not. We emphasize that both the consensus estimate and its estimated uncertainty depend on the assumed measurement error model, and we investigate measurement error model selection options for two examples.

DEVELOPMENT OF AN AUTOMATED TESTING SYSTEM FOR VERIFICATION AND VALIDATION OF NUCLEAR DATA AND SIMULATION CODE

Verification and validation (V&V) of nuclear data are critical to the accuracy of both stochastic and deterministic particle transport codes. To effectively test a set of nuclear data, the data must be applied to a wide variety of transport problems. Performing this task in a timely, efficient manner is tedious. The nuclear data team at Los Alamos National Laboratory in collaboration with the University of Florida has developed a methodology to automate the process of nuclear data V&V. This automated V&V process can efficiently test a number of data libraries using well-defined benchmark experiments, such as those in the International Criticality Safety Benchmark Experiment Project. The process is implemented through an integrated set of Python scripts. Material and geometry data are read from an existing medium or given directly by the user to generate a benchmark experiment template file. The user specifies the choice of benchmark templates, codes, and libraries to form a V“V project. The Python scripts automatically generate input decks for multiple transport codes, run and monitor individual jobs, and parse the relevant output. The output can then be used to generate reports directly or can be stored in a database for later analysis. This methodology eases the burden on the user by reducing the amount of time and effort required for obtaining and compiling calculation results. The resource savings by using this automated methodology could potentially be an enabling technology for more sophisticated data studies, such as nuclear data uncertainty quantification. Once deployed, this tool will allow the nuclear data community to more thoroughly test data libraries leading to higher-fidelity data in the future.

Release of ENDF/B-VIII.0-Based ACE Data Files

In February 2018, the National Nuclear Data Center released ENDF/B-VIII.0 [1] in the standard Evaluated Nuclear Data Format (ENDF) [2]. This represents the advances made in nuclear data since the release of ENDF/B-VII.1 [3] in 2011. The Nuclear Data Team at Los Alamos National Laboratory has processed the ENDF/BVIII.0 library and has made available a library of ACE data tables at several temperatures for each of the incident neutron ENDF/B files. The library is called Lib80x and is distributed from the website https://nucleardata.lanl.gov. The data was processed using NJOY2016.27 [4, 5]; a sample NJOY input deck is given in Appendix A. The release of the Lib80x library includes all the ENDF/B-VIII.0 incident neutron evaluations processed to the seven temperatures shown in Table 1. These are the same temperatures used for the ENDF/B-VII.1-based library, ENDF71x [6, 7].

Milestone 5431: Chi-Nu Measurements of Prompt Fission Neutron Spectra (PFNS)

This presentation offers an overview of the Chi-Nu project, its results, evaluation, and plans for FY17.