Publications

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Presented in this report is the description of a new method for neutron energy spectrum adjustment which uses a genetic algorithm to minimize the difference between calculated and measured reaction probabilities. The measured reaction probabilities are found using neutron activation analysis. The method adjusts a trial spectrum provided by the user which is typically calculated using a neutron transport code such as MCNP. Observed benefits of this method over currently existing methods include the reduction in unrealistic artefacts in the spectral shape as well as a reduced sensitivity to increases in the energy resolution of the derived spectrum. This report presents the adjustment results for various spectrum altering bucket environments in the central cavity of the Annular Core Research Reactor, as well as the adjustment results for the spectrum in the Sandia Pulse Reactor III. In each case, the results are compared to those generated using LSL-M2, which is a code commonly used for the purpose of spectrum adjustment. The genetic algorithm produces spectrum-averaged reaction probabilities with agreement to measured values, and comparable to those resulting from LSL-M2. The true benefit to this method, the reduction of shape artefacts in the spectrum, is difficult to quantify but can be clearly seen in the comparison of the final adjustments. Beyond these preliminary results, this report also gives a thorough description of the genetic algorithm and presents instructions for running the code using the graphical user interface. In its present state, the code does not provide uncertainties or correlations for the adjusted spectrum. This capability is currently being added, and will be presented in future work.

This document presents the facilit y - recommended characteri zation o f the neutron, prompt gamma - ray, and delayed gamma - ray radiation fields in the Annular Core Research Reactor ( ACRR ) for the cen tral cavity free - field environment with the 32 - inch pedestal at the core centerline. The designation for this environmen t is ACRR - FF - CC - 32 - cl. The neutron, prompt gamma - ray , and delayed gamma - ray energy spectra , uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma - ray fluence profiles within the experiment area of the cavity . Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples . Acknowledgements The authors wish to th ank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work . Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

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This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integral dosimetry measurements in the neutron field are reported.

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the 44 inch Lead-Boron (LB44) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity free-field reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.

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We describe a new method for neutron energy spectrum adjustment which uses a genetic algorithm to minimize the difference between calculated and measured reaction probabilities. The measured reaction probabilities are found using neutron activation analysis. The method adjusts a trial spectrum provided by the user which is typically calculated using a neutron transport code such as MCNP. Observed benefits of this method over currently existing methods include the reduction in unrealistic artifacts in the spectral shape as well as a reduced sensitivity to increases in the energy resolution of the derived spectrum. The method has thus far been used to perform spectrum adjustments on several spectrum-modifying environments in the central cavity of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, NM. Presented in this paper are the adjustment results for the polyethylene-lead-graphite (PLG) bucket environment along with a comparison to an adjustment obtained using the code LSL-M2, which uses a logarithmic least squares approach. The genetic algorithm produces spectrum-averaged reaction probabilities with agreement to measured values, and comparable to those resulting from LSL-M2. The true benefit to this method, the reduction of shape artifacts in the spectrum, is difficult to quantify but can be clearly seen in the comparison of the final adjustments.

The process nuclear criticality accident that occurred at the Mayak Production Association (Chelyabinsk-40) on January 2, 1958 involving a vessel of uranyl nitrate solution claimed the lives of three workers and left a fourth worker with continuing health problems. There are a myriad of uncertain parameters involved with this accident: What was the molarity of the solution? How much solution was in the vessel at the time of the accident? In what position was the vessel and the solution when it went critical? How important was the impact of reflection due to the workers and/or the floor? These uncertain parameters have made this accident particularly difficult to analyze in the past. This work aims to lower the uncertainty on some of these parameters. A most-probable solution composition is determined by comparing literature on the physical properties of uranyl nitrate solutions to those presented in LA-13638 [1], which describes the accident in question. Using this most-probable solution, the main contributions to the reactivity of the system and hence the eventual accident, are identified through Serpent 2 and OpenFOAM analyses. Serpent 2, a Monte Carlo software tool, is used to perform calculations of the reactivity effects of lowering the vessel toward the floor and the reactivity added by the close proximity of workers. OpenFOAM, a C++ partial differential equation solver toolkit, is used to simulate the fluid inside the vessel as the vessel is tipped. This is done by treating the solution and air inside the vessel as two incompressible, isothermal, and immiscible fluids using a volume of fluid (VoF) approach. The goal of this approach is simply to track the interface between the two fluids, and hence give an accurate description of the geometrical structure of the solution as the vessel is tipped. These two unique tools are then coupled to provide a time-dependent flow simulation to study the effect that the changing geometrical structure had on the criticality of the system, which is novel to the criticality safety field. This work provides a more accurate picture of the accident going forward. Key Words: Serpent 2, OpenFOAM, multi-physics, prompt neutron excursion, nuclear criticality safety accident, process condition change.

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