Publications

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High fidelity active dosimetry in the mixed neutron/gamma field of a research reactor is a very complex issue. For passive dosimetry applications, the use of activation foils addresses the neutron environment while the use of low neutron response CaF2:Mn thermoluminescent dosimeters (TLDs) addresses the gamma environment. While radiation-hardened diamond photoconducting detectors (PCD) have been developed that provide a very precise fast response (picosecond) dosimeter and can provide a time-dependent profile for the radiation environment, the mixed field response of the PCD is still uncertain and this interferes with the calibration of the PCD response. In order to address the research reactor experimenter's need for a dosimeter that reports silicon dose and dose rate at a test location during a pulsed reactor operation, a silicon calorimeter has been developed. This dosimeter can be used by itself to provide a dose in rad(Si) up to a point in a reactor pulsed operation, or, in conjunction with the diamond PCD, to provide a dose rate. This paper reports on the development, testing, and validation of this silicon calorimeter for applications in water-moderated research reactors. Copyright © 2006 by ASTM International.

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Thermoluminescent dosimeters (TLDs), particularly CaF2:Mn, are often used as photon dosimeters in mixed (n/γ) field environments. In these mixed field environments, it is desirable to separate the photon response of a dosimeter from the neutron response. For passive dosimeters that measure an integral response, such as TLDs, the separation of the two components must be performed by postexperiment analysis because the TLD reading system cannot distinguish between photon- and neutron-produced response. Using a model of an aluminum-equilibrated TLD-400 (CaF2:Mn) chip, a systematic effort has been made to analytically determine the various components that contribute to the neutron response of a TLD reading. The calculations were performed for five measured reactor neutron spectra and one theoretical thermal neutron spectrum. The five measured reactor spectra all have experimental values for aluminum-equilibrated TLD-400 chips. Calculations were used to determine the percentage of the total TLD response produced by neutron interactions in the TLD and aluminum equilibrator. These calculations will aid the Sandia National Laboratories-Radiation Metrology Laboratory (SNL-RML) in the interpretation of the uncertainty for TLD dosimetry measurements in the mixed field environments produced by SNL reactor facilities.

This paper defines a process for selecting dosimetry-quality cross sections. The recommended cross-section evaluation depends on screening high-quality evaluations with quantified uncertainties, down-selecting based on comparison to experiments in standard neutron fields, and consistency checking in reference neutron fields. This procedure is illustrated for the 23Na(n, γ)24 Na reaction.

The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current "state-of-the-art" in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two examples are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new "widget" to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation.

This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

Reactor neutron environments can be used to test/screen the sensitivity of unhardened commercial SRAMs to low-LET neutron-induced upset. Tests indicate both thermal/epithermal (< 1 keV) and fast neutrons can cause upsets in unhardened parts. Measured upset rates in reactor environments can be used to model the upset rate for arbitrary neutron spectra.

New dosimetry cross-section evaluations have been made available to the reactor community. Most dosimetry-quality evaluations include a section (File 33) that defines the uncertainty and covariance matrix for the dosimetry reaction cross section. This paper compares the latest computed cross-section activities for benchmark neutron fields with experimental data. Uncertainty data is usually reported with experimental measurements. This work also presents uncertainty data for the calculated activities. The calculated uncertainty values include a full uncertainty propagation using the cross-section evaluation, energy-dependent covariance data as well as the uncertainty attributed to the knowledge of the neutron spectrum.

The electronics radiation hardness-testing community uses the ASTM E722-93 Standard Practice to define the energy dependence of the nonionizing neutron damage to silicon semiconductors. This neutron displacement damage response function is defined to be equal to the silicon displacement kerma as calculated from the ORNL Si cross-section evaluation. Experimental work has shown that observed damage ratios at various test facilities agree with the defined response function to within 5%. Here, a covariance matrix for the silicon 1-MeV neutron displacement damage function is developed. This uncertainty data will support the electronic radiation hardness-testing community and will permit silicon displacement damage sensors to be used in least squares spectrum adjustment codes.

Fission foils are commonly used as dosimetry sensors. They play a very important role in neutron spectrum determinations. This paper provides a combination of experimental measurements and calculations to quantify the importance and synergy of several factors that affect the fission response of a dosimeter. Only when these effects are properly treated can fission dosimeters be used with sufficient fidelity.

Most neutron spectrum determination methodologies ignore self-shielding effects in dosimetry foils and treat covers with an exponential attenuation model. This work provides a quantitative analysis of the approximations in this approach. It also provides a methodology for improving the fidelity of the treatment of the dosimetry sensor response to a level consistent with the user`s spectrum characterization approach. A library of correction functions for the energy-dependent sensor response has been compiled that addresses dosimetry foils/configurations in use at the Sandia National Laboratories Radiation Metrology Laboratory.

Several new neutron cross section libraries, such as ENDF/ B-VI and IRDF-90, have recently been made available to the dosimetry community. Recommendations are made for the source selection of reaction cross sections that vary significantly among the libraries. In general, integral parameters from spectra obtained from unfold/adjustment codes using the new cross sections will not significantly change. A 61-reaction compendium of dosimetry cross sections drawn from existing evaluations has been compiled for use at the Sandia National Laboratories Radiation Metrology Laboratory. This dosimetry library (SNLRML) is recommended for use in spectrum determination with unfold/ adjustment methods. © 1992 IEEE

A 1-MeV neutron damage equivalence methodology and damage function have been developed for GaAs based on a recoil-energy dependent damage efficiency and the displacement kerma. This method, developed using life-time degradation in GaAs LEDs in a variety of neutron spectra, is also shown to be applicable to carrier removal. A validated methodology, such as this, is required to ensure and evaluate simulation fidelity in the neutron testing of GaAs semiconductors.

ENDF/B-VI cross sections were released to the testing community in January 1990. Work at Sandia National Laboratories, with pre-released versions of the new cross sections indicates that changes in the neutron-induced charged-particle reactions will significantly affect 14-MeV neutron dosimetry. Reactions that are important for fission reactor dosimetry were examined and most did not change significantly. 12 refs., 3 figs., 3 tabs.