Publications

Abstract not provided.

A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response.

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The IRDFF cross section library provides the highest fidelity cross section characterization and is the recommended data library to be used for dosimetry in support of reactor pressure vessel surveillance programs. In order to support this critical application, quantified validation evidence is required for the cross section library. Results are reported here on the use of various advanced approaches to uncertainty quantification using metrics relevant to spectrum characterization applications. The use of a quantified least squares approach, combining a consistent treatment of uncertainty from the spectral characterizations, the dosimetry cross sections, and measured activation products, is identified as one of the most sensitive metrics by which to report validation evidence. Using this metric the status of the validation of the IRDFF library was investigated. This analysis began with a consideration of the best characterized 252Cf spontaneous fission standard neutron benchmark field. Good validation evidence is found for 39 of the 79 IRDFF reactions. The 235U thermal fission reference neutron field was then investigated, and found to yield good validation evidence for an additional 10 of the IRDFF reactions. Extending the analysis further to include four different reactor-based reference neutron benchmark fields, ranging from fast burst reactors to well-moderated pool-type reactors, yielded good validation evidence for an additional 6 IRDFF reactions. In total, evidence is reported here for 55 of the 79 reactions in the IRDFF library.

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Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.

Abstract not provided.

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.

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.

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 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.

Abstract not provided.

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 polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-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 David Ames for his assistance in running MCNP on the Sandia parallel machines.

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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.

Abstract not provided.

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.

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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) for the 44-inch-long lead-boron bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-LB44-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra are presented as well as radial and axial neutron and gamma-ray flux 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 and steady-state operations are presented with conversion examples.

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.