The characterization of the uncertainty in radiation damage metrics presents many challenges. This paper examines the current approaches to characterizing radiation damage metrics such as hydrogen and helium gas production, material heating, trapped charge in microelectronics, and lattice displacement damage. Critical uncertainty aspects go beyond just the material cross sections and involve the consideration of energy-dependent cross reaction correlations, the recoil ion energy spectrum, and models used for the partitioning of the recoil ion energy into various forms of energy deposition. This paper starts with a review of terminology and then examines the current approaches in the characterization of uncertainty in radiation damage metrics for several applications. The major deficiencies in the uncertainty of the damage metric characterization are also identified.
The radiation effects community needs clear, well-documented, neutron energy-dependent responses that can be used in assessing radiation-induced material damage to GaAs semiconductors and for correlating observed radiation-induced changes in the GaAs electronic properties with computed damage metrics. In support of the objective, this document provides: a) a clearly defined set of relevant neutron response functions for use in dosimetry applications; b) clear mathematical expressions for the defined response functions; and c) updated quantitative values for the energy- dependent response functions that reflect the best current nuclear data and modelling. This document recaps the legacy response functions. It then surveys the latest nuclear data and updates the recommended response function to support current GaAs damage studies. A detailed tabulation for six of the energy-dependent response functions is provided in an Appendix.
The emerging use of the physics-based athermal recombination-corrected displacement per atom (arc-dpa) model for the displacement damage efficiency has motivated a re-evaluation of the historical empirically-derived GaAs damage response function with the purpose of highlighting needs for future analytical and experimental work. The 1-MeV neutron damage equivalence methodology used in the ASTM E-722 standard for GaAs has been re-evaluated using updated nuclear data. This yielded a higher fidelity representation of the GaAs displacement kerma and, through the use of the refined PKA recoil energy-dependent damage efficiency model, an updated 1-MeV(GaAs) displacement damage function. This re-evaluation included use of the Norgett-Robinson-Torrens (NRT) model for an updated threshold treatment, rather than the sharp-threshold Kinchin-Pease model used in the current ASTM standard. The underlying nuclear data evaluations have been updated to use the ENDF/VIII.0 75As and TENDL-2019 71Ga/69Ga evaluations. The displacement kerma and 1-MeV-equivalent damage responses were calculated using a modified NJOY-2016 code which allowed for refinements in some of the damage models. This paper shows that an updated displacement damage function, based upon the latest nuclear data, is consistent with the experimental data used to develop the current ASTM E-722 GaAs standard. Using a double ratio approach to compare the available experimental data with the calculated response, the average legacy double ratio was found to be 0.97±0.05 and the average updated double ratio was found to be 0.94 ±0.05.
The emerging use of the physics-based athermal recombination-corrected displacement per atom (arc-dpa) model for the displacement damage efficiency has motivated a re-evaluation of the historical empirically-derived GaAs damage response function with the purpose of highlighting needs for future analytical and experimental work. The 1-MeV neutron damage equivalence methodology used in the ASTM E-722 standard for GaAs has been re-evaluated using updated nuclear data. This yielded a higher fidelity representation of the GaAs displacement kerma and, through the use of the refined PKA recoil energy-dependent damage efficiency model, an updated 1-MeV(GaAs) displacement damage function. This re-evaluation included use of the Norgett-Robinson-Torrens (NRT) model for an updated threshold treatment, rather than the sharp-threshold Kinchin-Pease model used in the current ASTM standard. The underlying nuclear data evaluations have been updated to use the ENDF/VIII.0 75As and TENDL-2019 71Ga/69Ga evaluations. The displacement kerma and 1-MeV-equivalent damage responses were calculated using a modified NJOY-2016 code which allowed for refinements in some of the damage models. This paper shows that an updated displacement damage function, based upon the latest nuclear data, is consistent with the experimental data used to develop the current ASTM E-722 GaAs standard. Using a double ratio approach to compare the available experimental data with the calculated response, the average legacy double ratio was found to be 0.97±0.05 and the average updated double ratio was found to be 0.94 ±0.05.
This report provides basic background data on the Manipulate-2020 code. This code is used for processing and "manipulation" of nuclear data in support of radiation metrology applications. The code is made available on the open GitHub repository and is available to the general nuclear data community.
GaN has electronic properties that make it an excellent material for the next generation of power electronics; however, its radiation hardening still needs further understanding before it is used in radiation environments. In this work we explored the response of commercial InGaN LEDs to two different radiation environments: ion and gamma irradiations. For ion irradiations we performed two types of irradiations at the Ion Beam Lab (IBL) at Sandia National Laboratories (SNL): high energy and end of range (EOR) irradiations. For gamma irradiations we fielded devices at the gamma irradiation facility (GIF) at SNL. The response of the LEDs to radiation was investigated by IV, light output and light output vs frequency measurements. We found that dose levels up to 500 krads do not degrade the electrical properties of the devices and that devices exposed to ion irradiations exhibit a linear and non- linear dependence with fluence for two different ranges of fluence levels. We also performed current injection annealing studies to explore the annealing properties of InGaN LEDs.
Four D flip-flop (DFF) layouts were created from the same schematic in Sandia National Laboratories' CMOS7 silicon-on-insulator (SOI) process. Single-event upset (SEU) modeling and testing showed an improved response with the use of shallow (not fully bottomed) N-type metal-oxide-semiconductor field-effect transistors (NMOSFETs), extending the size of the drain implant and increasing the critical charge of the transmission gates in the circuit design and layout. This research also shows the importance of correctly modeling nodal capacitance, which is a major factor determining SEU critical charge. Accurate SEU models enable the understanding of the SEU vulnerabilities and how to make the design more robust.
Griffin, Patrick J.; Trkov, A.; Simakov, S.P.; Greenwood, L.R.; Zolotarev, K.I.; Capote, R.; Destouches, C.; Kahler, A.C.; Konno, C.; Kostal, M.; Aldama, D.L.; Chechev, V.; Majerle, M.; Malambu, E.; Ohta, M.; Pronyaev, V.G.; Yashima, H.; White, M.; Wagemans, J.; Vavtar, I.; Simeckova, E.; Radulovic, V.; Sato, S.
High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new IRDFF-II library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications.
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.
A general formulation of silicon damage metrics and associated energy-dependent response functions relevant to the radiation effects community is provided. Using this formulation, a rigorous quantitative treatment of the energy-dependent uncertainty contributors is performed. This resulted in the generation of a covariance matrix for the displacement kerma, the Norgett-Robinson-Torrens-based damage energy, and the 1-MeV(Si)-equivalent damage function. When a careful methodology is used to apply a reference 1-MeV damage value, the systematic uncertainty in the fast fission region is seen to be removed, and the uncertainty for integral metrics in broad-based fission-based neutron fields is demonstrated to be significantly reduced.
Here, a general formulation of silicon damage metrics and associated energy-dependent response functions relevant to the radiation effects community is provided. Using this formulation, a rigorous quantitative treatment of the energy-dependent uncertainty contributors is performed. This resulted in the generation of a covariance matrix for the displacement kerma, the NRT-based damage energy, and the 1-MeV(Si) equivalent damage function. Lastly, when a careful methodology is used to apply a reference 1-MeV damage value, the systematic uncertainty in the fast fission region is seen to be removed and the uncertainty for integral metrics in broad-based fission-based neutron fields is demonstrated to be significantly reduced.
The radiation effects community embraces the importance of quantifying uncertainty in model predictions and the importance of propagating this uncertainty into the integral metrics used to validate models, but they are not always aware of the importance of addressing the energy- and reaction-dependent correlations in the underlying uncertainty contributors. This paper presents a rigorous high-fidelity Total Monte Carlo approach that addresses the correlation in the underlying uncertainty components and quantifies the role of both energy and reaction-dependent correlations in a sample application that addresses the damage metrics relevant to silicon semiconductors.
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.
A rigorous treatment of the uncertainty in the underlying nuclear data on silicon displacement damage metrics is presented. The uncertainty in the cross sections and recoil atom spectra are propagated into the energy-dependent uncertainty contribution in the silicon displacement kerma and damage energy using a Total Monte Carlo treatment. An energy-dependent covariance matrix is used to characterize the resulting uncertainty. A strong correlation between different reaction channels is observed in the high energy neutron contributions to the displacement damage metrics which supports the necessity of using a Monte Carlo based method to address the nonlinear nature of the uncertainty propagation.
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.
The effect of uncertainty in the energy partition function on the silicon displacement damage metric is presented. Through the use of a Total Monte Carlo approach, the effect of uncertainty in the underlying electronic and nuclear ion interaction potentials, which are used to define the damage partition function, is propagated into an uncertainty in the silicon damage metric. This uncertainty is expressed as an energy-dependent covariance matrix which permits this uncertainty component to be combined with other uncertainty components, e.g. uncertainty due to the knowledge of the nuclear interaction data or to the treatment of the damage in the threshold displacement region. This approach provides a rigorous treatment of uncertainty due to the damage metric which can then be propagated in uncertainty estimates for various applications, e.g. when examining damage equivalence between different neutron sources. A strong energy-dependent correlation is found in this uncertainty component.
This report provides a set of consistent definitions for radiation damage metrics relevant to the modeling of displacement damage in materials. The limitations/approximations built into the various metrics are discussed, as are the intended applications that gave rise to community use of the damage metrics. Numerical tabulations are provided, based on the latest nuclear data, for recommended values of the neutron displacement kerma factor and various NRT-based damage energy metrics in silicon.
Griffin, Patrick J.; Simakov, Stanislav; Capote, Roberto; Greenwood, Lawrence; Kahler, Albert; Trkov, Andrej; Zolotarev, Konstantin; Pronyaev, Vladimir
The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.
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.
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 central cavity free-field environment with the 32-inch pedestal at the core centerline. The designation for this environment 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.
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.
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.
In this study, the dosimetry community has a history of using spectral indices to support neutron spectrum characterization and cross section validation efforts. An important aspect to this type of analysis is the proper consideration of the contribution of the spectrum uncertainty to the total uncertainty in calculated spectral indices (SIs). This study identifies deficiencies in the traditional treatment of the SI uncertainty, provides simple bounds to the spectral component in the SI uncertainty estimates, verifies that these estimates are reflected in actual applications, details a methodology that rigorously captures the spectral contribution to the uncertainty in the SI, and provides quantified examples that demonstrate the importance of the proper treatment the spectral contribution to the uncertainty in the SI.
This report provides a set of consistent definitions for metrics relevant to the modeling of displacement damage in materials. The limitations/approximations built into the various metrics are discussed as is the intended application that gave rise to community use of the metric. Recommended sources for numerical tabulations of the neutron displacement kerma and the charged particle non-ionizing energy loss in some important semiconductor materials are also 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 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.
A fast neutron detector is being developed to measure the cosmic ray neutron flux in order to measure soil moisture. Soil that is saturated with water has an enhanced ability to moderate fast neutrons, removing them from the backscatter spectrum. The detector is a two-element, liquid scintillator detector. The choice of liquid scintillator allows rejection of gamma background contamination from the desired neutron signal. This enhances the ability to reconstruct the energy and direction of a coincident neutron event. The ability to image on an event-by-event basis allows the detector to selectively scan the neutron flux as a function of distance from the detector. Calibrations, simulations, and optimization have been completed to understand the detector response to neutron sources at variable distances and directions. This has been applied to laboratory background measurements in preparation for outdoor field tests.
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.
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.
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.
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.
Various metrics are formulated for the uncertainty of calculated neutron activities for dosimetry reactions. The correlations between the uncertainty metrics are examined. The uncertainty data are presented for the dosimetry reactions and can be used to guide the selection of sensors used in spectrum determinations.
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.
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.