Publications

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Neutron dosimetry monitors should be used during all irradiations in the Annular Core Research Reactor. This report provides the recommended conversion factors that should be used to translate the monitor dosimeter read-outs into the damage metrics that are typically used by experimenters to assess the results of their experiment. These conversion factors are based upon the use of the latest least-squares adjusted neutron spectrum determination to describe the Sandia National Laboratories reference neutron fields and the latest International Atomic Energy Agency recommended dosimetry cross sections to capture the response of the dosimeter. The resulting conversion factors are built into the dosimetry results routinely provided by the Radiation Metrology Laboratory.

This document presents the process for a new method developed for the characterization of the delayed gamma-ray radiation fields in pulse reactors like the Annular Core Research Reactor (ACRR) and the Fueled Ring External Cavity (FREC-II). The environments used to test this method in the ACRR were FF, LB44, PLG and CdPoly, and the environments used in the FREC-II were FF with rods-down, FF with rods-up, CdPoly with rods-down and CdPoly with rods-up. All environment configurations used the same fission product gamma-ray source energy spectrum. This method required the fission sites located in the MCNP KCODE source tapes. A FORTRAN script was written to translate and extract the coordinates for the fission sites. The 10K fission sites were then input it into an MCNP SOURCE mode script. Using a MATLAB script, a parametric analysis was done, and it helped determine that 10K fission sites are an appropriate number of coordinates to converge to the correct answer. The method gave excellent results and was tested in the ACRR, FREC-II and White Sands Missile Range (WSMR). This method can be applied to other pulse research reactors as well.

An overview of experimental and computational studies of prompt secondary gamma production and transport, executed under the auspices of the Readiness in Technical Base and Facilities (RTBF) program, is presented. Relevant experiments at the Annular Core Research Reactor (ACRR) were conducted in the FY2012 -- FY2014 timeframe and pertain to the performance of various elemental calorimeters and the analytic fractionation of dose contributions to the calorimeter discs. In particular, the influence of the choice of prompt capture gamma production databases on the computed disc heating factors is discussed. Finally, the results of a polyurethane foam moderation experiment are detailed.

Silicon calorimeters have been used for active radiation dosimetry in the central cavity of the Annular Core Research Reactor (ACRR) for over a decade. Recently, there has been interest in using other materials for calorimetry to accurately measure the prompt gamma-ray energy deposition in the mixed neutron and gamma-ray environment. The calorimeters used in the ACRR use a thermocouple (TC) to measure the change in temperature of specific materials in the radiation environment. The temperature change is related to the instantaneous dose received by the material in a pulse-transient operation. SOLIDWORKS Simulation and ANSYS Mechanical were used to model the calorimeter and analyze the thermal behavior under pulse-transient conditions. This report compares the results from modeling to experimental results for selected calorimeter materials and radiation environments. These materials include bismuth, tin, zirconium, and silicon. Calorimeters assembled with each material were irradiated in the ACRR central cavity in the free- field, LB44, CdPoly, and PLG radiation environments. The neutronics code Monte-Carlo N- Particle (MCNP) was used to calculate the neutron and gamma-ray response of the calorimeter materials at the experimental locations in the central cavity. Different response tallies were used and found to give different results for the gamma-ray energy deposition. It was determined that performing the neutron/gamma-ray/electron transport in MCNP using the *F8 electron tally gave the overall best agreement with the experimental results. The *F8 tally, however, is much more computationally intensive than the neutron/gamma-ray transport calculations. Also, this report contains parametric analyses that examine the ways to improve the current design of the calorimeters. One finding from the parametric analysis was that the TC should be placed closer to the outer radius of the disks to obtain a measurement closer to the maximum temperature of the disk. Also, the parametric analysis showed that the most dominant mechanism of heat loss in the calorimeters is conduction through the alumina posts. In future designs, the conduction should be minimized to reduce the effect of heat loss on the measurements.

Slides for AWE to be used in report, poster, or publication as UUR. Slides represent work performed at the ACRR in support of the DUSTER program.

A bipolar-transistor-based sensor technique has been used to compare silicon displacement damage from known and unknown neutron energy spectra generated in nuclear reactor and high-energy-density physics environments. The technique has been shown to yield 1-MeV(Si) equivalent neutron fluence measurements comparable to traditional neutron activation dosimetry. This paper significantly extends previous results by evaluating three types of bipolar devices utilized as displacement damage sensors at a nuclear research reactor and at a Pelletron particle accelerator. Ionizing dose effects are compensated for via comparisons with 10-keV X-ray and/or cobalt-60 gamma ray irradiations. Nonionizing energy loss calculations adequately approximate the correlations between particle and device responses and provide evidence for the use of one particle type to screen the sensitivity of the other.

The process of determining the uncertainty in the neutron fluence from the measured activity of a dosimetry monitor is reviewed and the importance of treating the energy-dependent correlation is illustrated using several representative neutron fields. The process of determining the uncertainty in the neutron fluence when a transfer calibration is used is then detailed. The conversion factor, when a transfer calibration is used, has a term that has an integral over the cross section appearing in both the numerator and the denominator. This term introduces a nonlinear dependence on the cross section within the conversion factor and an explicit correlation between the terms appearing in the numerator and denominator of the conversion factor. A method for rigorously treating this nonlinear uncertainty propagation is presented. This method is based upon utilizing the covariance matrix for the cross section and utilizing a statistical sampling approach based on a Cholesky transformation of this covariance matrix. This methodology is then applied to the determination of the uncertainty from a transfer calibration for a set of nine neutron spectra based upon using the 32S(n,p)32P reaction and a transfer calibration in a 2 5 2Cf standard benchmark neutron field. A very strong correlation is found in the cross-section terms as they appear in the numerator and in the denominator. When a rigorous treatment is used to propagate the uncertainty due to the cross section for the dosimetry monitor, the uncertainty in the conversion factor is reduced by a factor of more than ten times from a worst-case approach that treats the uncertainty components in the numerator and denominator as uncorrelated. This ten times difference is also seen when the comparison is made between a rigorous treatment and a treatment of the cross-section contributions where the numerator and denominator are treated as uncorrelated (i.e., when compared to a root-mean-square approach).

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories provides experimenters with a unique platform for irradiations. Its central cavity is wide enough to accommodate spectrum-modifying materials, commonly referred to as buckets. The addition of hydrogenous moderators, such as polyethylene or water, can cause considerable thermalization of the free field neutron spectrum. Conversely, thick annular regions of strong, thermal absorbers, such as boron or cadmium, create a faster neutron spectrum inside. Similarly, the gamma-ray fluence can be attenuated by adding high-Z materials or enhanced through radiative capture in cadmium or gadolinium. Novel configurations of buckets allow simultaneous neutron energy spectrum modification and gamma-ray attenuation. As such, different radiation environments can exist at ACRR's core centerline. Recent efforts have produced detailed characterizations of several neutron- and gamma-ray spectrum-modifying buckets for the ACRR central cavity, including: the free field; the 44-in.-tall lead-boron carbide bucket (fast neutron, attenuated photon); the polyethylene-lead-graphite bucket (thermalized neutrons, attenuated photon); and the Cd-Poly bucket (cadmium polyethylene lined bucket used to enhance photon production). Dedicated opportunities to perform multiple characterizations occurred somewhat infrequently, which afforded the authors the ability to hone techniques for performing these tests. Each neutron spectrum characterization generally followed both ASTM E720, Standard Guide for Selection and Use of Neutron Sensors for Determining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics, and ASTM E721, Standard Guide for Determining Neutron Energy Spectra from Neutron Sensors for Radiation-Hardness Testing of Electronics. This paper presents some practical lessons learned throughout these characterizations-both experimental and computational.

Presented in this report is the description of a new method for neutron energy spectrum adjustment that uses a genetic algorithm to minimize the difference between calculated and measured reaction probabilities. The reaction probability is the integral over all energies of the product of the microscopic reaction cross section with the neutron fluence. The measured reaction probabilities are found using neutron activation analysis. The method adjusts a trial spectrum provided by the user that typically is calculated using a neutron transport code such as Monte Carlo N-Particle. Observed benefits of this method over currently existing methods include: (a) the reduction in unrealistic artifacts in the spectral shape when compared to iterative unfold approaches such as are used in the SAND-II code or to least squares approaches when an accurate prior spectrum covariance is not available; and (b) 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. In each case, the results are compared to those generated using LSL-M2, which is a code commonly used for spectrum adjustment. The genetic algorithm produces spectrum-averaged reaction probabilities comparable to those resulting from LSL-M2. The splicing of local segments of the a priori spectrum, which is part of the genetic algorithm, permits the resulting spectrum adjustment to avoid introducing severe narrow energy-width shape artifacts without the requirement of a covariance matrix for the prior spectrum.

The Sandia National Laboratories' Annular Core Research Reactor (ACRR) is a unique pool-type research reactor that can pulse up to 300 MJ in energy. The ACRR maintains a dry, 9-in. (22.9 cm) diameter central cavity that extends through the center of the core region and allows for experiment irradiations at the peak neutron flux of the core. An epithermal/fast neutron flux exists in the cavity that allows the neutron energy spectrum to be modified to meet the requirements of the experimenter. Using a moderating material such as water or polyethylene in an annular geometry in the cavity allows a greater thermal neutron energy spectrum to be attained. Using a thermal neutron-absorbing material such as boron carbide or cadmium in an annular geometry in the cavity allows for a more epithermal-fast neutron energy spectrum. The gamma-ray fluence can be decreased by adding a high-Z material such as lead in an annular geometry. The gamma-ray fluence can be enhanced by adding a radiative capture material such as cadmium or gadolinium to a moderating material. Both neutron energy spectrum modification and gamma-ray attenuation/enhancement can be attained simultaneously. Different types of spectrum-modifying "buckets" are currently available for use by experimenters, and others can be custom designed and fabricated. This paper presents the results from the neutron and prompt gamma-ray characterization work for several of the environments in the ACRR central cavity, including the free field, polyethylene-lead-graphite, lead-boron-44 in., and cadmium-polyethylene bucket environments, and for the ACRR-Fueled Ring External Cavity-ll. These environments represent typical neutron and gamma-ray spectrum modifications that can be attained at the ACRR.

The objective of this report is to determine the feasibility of a combined pulsed - power accelerator machine, similar to HERMES - III, with the Annular Core Research Reactor (ACRR) Fueled - Ring External Cavity (FREC - II) in a new facility. The document is conceptual in nature, and includes some neutronic analysis that i llustrates that that the physics of such a concept would be feasible. There would still be many engineering design considerations and issues that would need to be investigated in order to determine the true viability of such a concept. This report does n ot address engineering design details, the cost of such a facility, or what would be required to develop the safety authorization of the concept. The radiation requirements for the "on - target" gamma - ray dose and dose rate are not addressed in this report . It is assumed that if the same general on - target specifications for a HERMES - III type machine could be met with the proposed concept, that the machine would b e considered highly useful as a radiation effects sciences platform. In general, the combined accelerator/ACRR reactor concept can be shown to be feasible with no major issues that would preclude the usefulness of such a facility. The new facility would provide a capability that currently does not exist in the radiation testing complex.

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This document presents the facility - recommended characterization of the neutron, prompt gamma - ray, and delayed gamma - ray radiation fields in the Annular Core Research Reactor ( ACRR ) Fueled - Ring External Cavity II (FREC - II) for the free - field environment at the core centerline. The designation for this environment is ACRR - FRECII - FF - 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.

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The capability to characterize the neutron energy spectrum and fluence received by a test object is crucial to understanding the damage effects observed in electronic components. For nuclear research reactors and high energy density physics facilities this can pose exceptional challenges, especially with low level neutron fluences. An ASTM test method for characterizing neutron environments utilizes the 2N2222A transistor as a 1-MeV equivalent neutron fluence sensor and is applicable for environments with 1 × 1012 - 1 × 1014 1 -MeV(Si)-Eqv.-n/cm2. In this work we seek to extend the range of this test method to lower fluence environments utilizing the 2N1486 transistor. The 2N1486 is shown to be an effective neutron displacement damage sensor as low as 1 × 1010 1-MeV(Si)-Eqv.-n/cm2.

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the cadmium-polyethylene (CdPoly) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-CdPoly-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 bucket. 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 thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to Drew Tonigan for helping field the activation experiments in ACRR, David Samuel for helping to finalize the drawings and get the parts fabricated, and Elliot Pelfrey for preparing the active dosimetry plots.

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