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DOE OSTI2021

Neutron Dosimetry for the Colorado School of Mines (CSM 16-10584) Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the Colorado School of Mines (CSM 16-10584) experiment which was conducted in position B5 of the ATR. Three other irradiations included in the scope of work are reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the CSM assemblies at INL prior to irradiation. Following irradiation, the capsules were returned to PNNL for analysis. The neutron dosimetry capsules were opened, the flux wires were removed for gamma analysis, and the measured activities were used to determine the activation rates for various activation products. Following suitable corrections, the measured activation rates were used to adjust calculated neutron spectra at 12 fluence monitor locations. The adjusted neutron spectra were then used to determine displacement per atom (dpa) and gas production for irradiated materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Westcott g factors extended to arbitrary neutron energy spectra

Westcott 𝑔 factors are used in Neutron Activation Analysis (NAA) and Prompt Gamma-ray Activation Analysis (PGAA) to evaluate the impact of non-1∕𝑣 behavior in the neutron-capture cross sections of certain nuclei on activation product yields. This non-1∕𝑣 behavior arises from the presence of neutron resonances in the neutron- capture cross sections that overlap with the source neutron spectrum at low (< 5 eV) energies. Historically, Westcott 𝑔 factors that have been cataloged for NAA and PGAA applications are the result of calculations that assume a Maxwellian neutron flux distribution with a given temperature. In this work, we use this approach with updated neutron-capture cross sections from the Evaluated Nuclear Data File, version VIII.1 (ENDF/B-VIII.1) to tabulate Westcott 𝑔 factor values for a broad range of Maxwellian distribution temperatures, comparing the results against currently-available 𝑔 factors from International Atomic Energy Agency tables and other sources. Here, it was discovered during this analysis that the use of guided thermal and cold-neutron beams at certain facilities necessitates an approach for evaluating Westcott 𝑔 factors based on arbitrary non-Maxwellian spectra. In this paper, we present an approach for calculating 𝑔 factors with user-specified neutron spectra, and we demonstrate these methods to obtain Westcott 𝑔-factors for guided- and cold-neutron beams at the Budapest Research Reactor and the Forschungsreaktor München II reactor. As part of this work, open-source software has been developed that can be used to perform these calculations for applications in PGAA and NAA experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2011

Standardizing Activation Analysis: New Software for Photon Activation Analysis

Photon Activation Analysis (PAA) of environmental, archaeological and industrial samples requires extensive data analysis that is susceptible to error. For the purpose of saving time, manpower and minimizing error, a computer program was designed, built and implemented using SQL, Access 2007 and asp.net technology to automate this process. Based on the peak information of the spectrum and assisted by its PAA library, the program automatically identifies elements in the samples and calculates their concentrations and respective uncertainties. The software also could be operated in browser/server mode, which gives the possibility to use it anywhere the internet is accessible. By switching the nuclide library and the related formula behind, the new software can be easily expanded to neutron activation analysis (NAA), charged particle activation analysis (CPAA) or proton-induced X-ray emission (PIXE). Implementation of this would standardize the analysis of nuclear activation data. Results from this software were compared to standard PAA analysis with excellent agreement. With minimum input from the user, the software has proven to be fast, user-friendly and reliable.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2013

Uranium and plutonium prompt-fission-neutron energy spectrum (PFNS) from the analysis of NTS NUEX data

The neutron experiment (NUEX) was a common diagnostic on nuclear device tests conducted at the Nevada Test Site (NTS). In these experiments neutrons from a device pass up a collimated line of site, and in the case of a Faraday cup (FC) NUEX, the neutrons pass through a thin CH 2 foil. Some of these neutrons interact with the nuclei in the foil, generating light charged particles (predominately protons) which are collected in a Faraday cup. The time dependence of the Faraday cup current is a measure of the energy spectrum of the neutrons that leak from the device. With good device models and accurate neutron-transport codes, the leakage spectrum can be converted into a prompt fast-neutron-induced fission-neutron energy spectrum (PFNS) from ~1 to 11 MeV. This has been done for two of our events containing a plutonium primary, where the NUEX data were of a particularly high quality, and one event containing a uranium primary. The fission-neutrons in the device were produced by fission events induced by neutrons over a broad range of energies. We have listed the inferred 1.5 MeV n + 239 Pu fission-neutron spectra in Tables 1 to 4 for outgoing neutron energies from 1.5 to ~10.5 MeV, in 1-MeV steps. The uranium device contained a larger high explosive charge. This limited the extraction of the uranium PFNS spectrum to upper neutron energy 9.5 MeV. The listed values represent the fission-neutron emission probability at the quoted outgoing neutron energies and are not the integrals over 1-MeV wide bins. The quoted relative emission probabilities are all relative to the probability of emitting 1.5 MeV neutrons. The presence of the high explosive charge surrounding the fissile material made estimates of the lower energy PFNS (below 1 MeV) problematic. To obtain estimates of the absolute emission probabilities, the low-energy portion of the PFNS was assumed to be as calculated by the Los Alamos (fission) model (LAM). The three NUEX inferred PFNS are labeled Pu-NUEX-1; Pu-NUEX-2; and U-NUEX-1. Pu-NUEX-1 was from an event near the end of US nuclear testing and its data quality was higher than the data associated with the earlier events used to obtain Pu-NUEX-2 and U-NUEX-1.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Improved field assessments of chemicals and explosives using high-resolution gamma-ray spectroscopy with a tagged neutron interrogation system

A fi eld-portable tagged neutron interrogation system is being developed for detection of explosives and chemical warfare agents by prompt gamma-ray neutron activation analysis. The system combines an associated particle imaging deuterium-tritium neutron generator with a high-purity germanium gamma-ray detector. Fine spatial resolution neutron and alpha detectors enable precise mapping of characteristic gamma-ray emission distributions onto high-clarity neutron radiographic images. A narrow coincidence gate associated with each tagged neutron ensures a large signal-to-background ratio, thus signifi cantly improving sensitivity to the chemical compositions of the items being assessed. Recent experimental results recorded from test objects involving mock explosives and other chemicals are presented.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2025

Improved field assessments of chemicals and explosives using high-resolution gamma-ray spectroscopy with a tagged neutron interrogation system

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2014

In-situ neutron diffraction of LaCoO₃ perovskite under uniaxial compression. I. Crystal structure analysis and texture development

The dynamics of texture formation, changes in crystal structure, and stress accommodation mechanisms have been studied in perovskite-type R3⁻c rhombohedral LaCoO₃ during uniaxial compression using in-situ neutron diffraction. The in-situ neutron diffraction revealed the complex crystallographic changes causing the texture formation and significant straining along certain crystallographic directions during compression, which are responsible for the appearance of hysteresis and non-linear ferroelastic deformation in the LaCoO₃ perovskite. The irreversible strain after the first loading was connected with the appearance of non-recoverable changes in the intensity ratio of certain crystallographic peaks, causing non-reversible texture formation. However, in the second loading/unloading cycle, the hysteresis loop was closed and no further irrecoverable strain appeared after deformation. The significant texture formation is responsible for an increase in the Young's modulus of LaCoO₃ at high compressive stresses, ranging from 76 GPa at the very beginning of the loading to 194 GPa at 900 MPa at the beginning of the unloading curve.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2014

Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO 2 /zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2023

R-Matrix Analysis of the Neutron-Induced Cross Sections on 143 Nd and 147,149 Sm Measured at LANSCE [Slides]

This presentation covers how an established used of R-Matrix codes is being implemented at LANSCE. This presentation displays how there are unique features accessible with detectors at LANSCE that can be used to perform more complete R-Matrix analysis. Additionally discussed is how spin separation is a powerful technique that can be applied in DANCE measurements. Unique method to measure transmission with DICER are discussed. Finally, transmission and capture data are being analyzed in this work. The study can be extended to other channels. The ongoing analysis of the 143 Nd data is presented in this work. The same method is being applied to analyze the 147,149 Sm data. The 143 Nd and 147,149 Sm work started in FY-23 funded by NCSP.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

Neutron Dosimetry for the Boise State University JW 15-8242 Irradiation in ATR (Rev.1)

PNNL project 79550 provides for the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00269673and Statement of Work No. 19704, Rev. 0, PNNL Analysis of NSUF Flux Capsules. This report is for the Boise State University (BSU) JW 15-8242 irradiations which were conducted in positions A6, A7 and A8 of the ATR. We previously reported data for this experiment after about 1 dpa in a report issued in January 2021 (1). This report is for longer irradiations to about 3 dpa. Other experiments included in the statement of work for this project will be reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the BSU assemblies prior to irradiation. Following irradiation, the capsules were returned to PNNL for analysis. The neutron dosimetry capsules were opened, the flux wires were removed for gamma or x-ray analysis, and the measured activities were used to determine the activation rates for various activation products. Following suitable corrections, the measured activation rates were used to adjust calculated neutron spectra at each fluence monitor location. The adjusted neutron spectra were then used to determine displacement per atom (dpa) and gas production for irradiated materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Berkeley Low Background Facility

The Berkeley Low Background Facility (BLBF) at Lawrence Berkeley National Laboratory (LBNL) in Berkeley, California provides low background gamma spectroscopy services to a wide array of experiments and projects. The analysis of samples takes place within two unique facilities; locally within a carefully-constructed, low background laboratory on the surface at LBNL and at the Sanford Underground Research Facility (SURF) in Lead, SD. These facilities provide a variety of gamma spectroscopy services to low background experiments primarily in the form of passive material screening for primordial radioisotopes (U, Th, K) or common cosmogenic/anthropogenic products; active screening via neutron activation analysis for U,Th, and K as well as a variety of stable isotopes; and neutron flux/beam characterization measurements through the use of monitors. A general overview of the facilities, services, and sensitivities will be presented. Recent activities and upgrades will also be described including an overview of the recently installed counting system at SURF (recently relocated from Oroville, CA in 2014), the installation of a second underground counting station at SURF in 2015, and future plans. The BLBF is open to any users for counting services or collaboration on a wide variety of experiments and projects.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2014

SU-E-T-495: Neutron Induced Electronics Failure Rate Analysis for a Single Room Proton Accelerator

Purpose: To determine the failure rate as a function of neutron dose of the range modulator's servo motor controller system (SMCS) while shielded with Borated Polyethylene (BPE) and unshielded in a single room proton accelerator. Methods: Two experimental setups were constructed using two servo motor controllers and two motors. Each SMCS was then placed 30 cm from the end of the plugged proton accelerator applicator. The motor was then turned on and observed from outside of the vault while being irradiated to known neutron doses determined from bubble detector measurements. Anytime the motor deviated from the programmed motion a failure was recorded along with the delivered dose. The experiment was repeated using 9 cm of BPE shielding surrounding the SMCS. Results: Ten SMCS failures were recorded in each experiment. The dose per monitor unit for the unshielded SMCS was 0.0211 mSv/MU and 0.0144 mSv/MU for the shielded SMCS. The mean dose to produce a failure for the unshielded SMCS was 63.5 ± 58.3 mSv versus 17.0 ±12.2 mSv for the shielded. The mean number of MUs between failures were 2297 ± 1891 MU for the unshielded SMCS and 2122 ± 1523 MU for the shielded. A Wilcoxon Signed Ranked test showed the dose between failures were significantly different (P value = 0.044) while the number of MUs between failures were not (P value = 1.000). Statistical analysis determined a SMCS neutron dose of 5.3 mSv produces a 5% chance of failure. Depending on the workload and location of the SMCS, this failure rate could impede clinical workflow. Conclusion: BPE shielding was shown to not reduce the average failure of the SMCS and relocation of the system outside of the accelerator vault was required to lower the failure rate enough to avoid impeding clinical work flow.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2012

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Non-destructive Material Characterization of Radioactive Waste Packages with QUANTOM{sup R} - 20130

During the last decades, the nuclear and non-nuclear industry has produced a considerable amount of low and intermediate level radioactive wastes. In many countries, it is foreseen to dispose of these waste packages safely in a final repository under strict waste acceptance requirements (e.g. the radiological and material characterization) defined by national licensing and supervisory authorities. The material characterization of waste packages can be performed on the basis of existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. We offer here an innovative non-destructive technology called QUANTOM{sup R} (Quantitative Analysis of Toxic and nontoxic Materials) based on prompt and delayed gamma neutron activation analysis (P and DGNAA). This technology is able to identify, quantify and thus verify the amount of hazardous and non-hazardous substances in 200-l radioactive drums, which is required for a final disposal characterization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Accelerating Data Acquisition, Reduction, and Analysis at the Spallation Neutron Source

ORNL operates the world's brightest neutron source, the Spallation Neutron Source (SNS). Funded by the US DOE Office of Basic Energy Science, this national user facility hosts hundreds of scientists from around the world, providing a platform to enable break-through research in materials science, sustainable energy, and basic science. While the SNS provides scientists with advanced experimental instruments, the deluge of data generated from these instruments represents both a big data challenge and a big data opportunity. For example, instruments at the SNS can now generate multiple millions of neutron events per second providing unprecedented experiment fidelity but leaving the user with a dataset that cannot be processed and analyzed in a timely fashion using legacy techniques. To address this big data challenge, ORNL has developed a near real-time streaming data reduction and analysis infrastructure. The Accelerating Data Acquisition, Reduction, and Analysis (ADARA) system provides a live streaming data infrastructure based on a high-performance publish subscribe system, in situ data reduction, visualization, and analysis tools, and integration with a high-performance computing and data storage infrastructure. ADARA allows users of the SNS instruments to analyze their experiment as it is run and make changes to the experiment in real-time and visualize the results of these changes. In this paper we describe ADARA, provide a high-level architectural overview of the system, and present a set of use-cases and real-world demonstrations of the technology.

Campbell, Stuart I↗

DOE OSTI2022

Small angle neutron scattering and lipidomic analysis of a native, trimeric PSI-SMALP from a thermophilic cyanobacteria

The use of styrene-maleic acid copolymers (SMAs) to produce membrane protein-containing nanodiscs without the initial detergent isolation has gained significant interest over the last decade. We have previously shown that a Photosystem I SMALP from the thermophilic cyanobacterium, Thermosynechococcus elongatus (PSI-SMALP), has much more rapid energy transfer and charge separation in vitro than detergent isolated PSI complexes. In this study, we have utilized small-angle neutron scattering (SANS) to better understand the geometry of these SMALPs. These techniques allow us to investigate the size and shape of these particles in their fully solvated state. Further, the particle's proteolipid core and detergent shell or copolymer belt can be interrogated separately using contrast variation, a capability unique to SANS. Here we report the dimensions of the Thermosynechococcus elongatus PSI-SMALP containing a PSI trimer. At ~1.5 MDa, PSI-SMALP is the largest SMALP to be isolated; our lipidomic analysis indicates it contains ~1300 lipids/per trimeric particle, >40-fold more than the PSI-DDM particle and > 100 fold more than identified in the 1JB0 crystal structure. Interestingly, the lipid composition to the PSI trimer in the PSI-SMALP differs significantly from bulk thylakoid composition, being enriched ~50 % in the anionic sulfolipid, SQDG. Finally, utilizing the contrast match point for the SMA 1440 copolymer, we also can observe the ~1 nm SMA copolymer belt surrounding this SMALP for the first time, consistent with most models of SMA organization.

59 BASIC BIOLOGICAL SCIENCES↗

NASA NTRS1993

Trace-element composition of Chicxulub crater melt rock, K/T tektites and Yucatan basement

The Cretaceous/Tertiary (K/T) boundary Chicxulub impact is the best preserved large impact in the geologic record. The Chicxulub crater has been buried with no apparent erosion of its intracrater deposits, and its ejecta blanket is known and is well preserved at hundreds of localities globally. Although most of the molten material ejected from the crater has been largely altered, a few localities still preserve tektite glass. Availability of intra- and extracrater impact products as well as plausible matches to the targeted rocks allows the comparison of compositions of the different classes of impact products to those of the impacted lithologies. Determination of trace-element compositions of the K/T tektites, Chicxulub melt rock, and the targeted Yucatan silicate basement and carbonate/evaporite lithologies have been made using instrumental neutron activation analysis (INAA) and inductively coupled plasma mass spectrometry (ICP-MS). Some sample splits were studied with both techniques to ensure that inter-laboratory variation was not significant or could be corrected. The concentration of a few major and minor elements was also checked against microprobe results. Radiochemical neutron activation analysis (RNAA) was used to determine Ir abundances in some samples.

Hildebrand, A. R.↗

DOE OSTI2021

Fatigue analysis of the Spallation Neutron Source 2 MW target design

Upgrades to the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory are under way with two major projects. The first project is the Proton Power Upgrade (PPU), which will double the power capacity of the SNS accelerator system to enable the operation of a future second neutron source target station—the second project. PPU also will increase power to the SNS first target station – which currently operates at 1.4 MW with 1.0 GeV protons – to 2 MW with 1.3 GeV protons. Final design of the PPU 2 MW target module is finished, including all necessary design and analysis calculations. The fatigue evaluation described herein is an assessment of the fitness of the PPU 2 MW target mercury vessel to resist failure from cyclic loading from the pulsed 60 Hz proton beam and thermal changes from disrupted operation. In addition to the reliability predictions, the method, detail, and depth of this fatigue evaluation are superior to those of past target design assessments. Furthermore, the thermal fatigue life is predicted to exceed 3125 thermal cycles per 1250 h of operation for all load cases. The design also meets the (relative) fatigue design goal for combined thermal and pulse loading of greater than 0.5 times the minimum fatigue life calculated for a jet-flow target design operating at 1.4 MW—the most robust design to date in terms of resistance to fatigue failure. From the onset of the PPU, it has been known that pulse loading requires effective application of helium gas injection into the target mercury to reduce beam pressure loading. For the PPU 2 MW target, none of the pulse load cases requires a maximum strain reduction from gas injection of more than 50% (for base material) to meet the fatigue design goal. This level of reduction is achievable from gas injection, according to historical strain measurements from 1.4 MW target operation. Overall, the PPU 2 MW target design is predicted to have superior resistance to fatigue failure compared with past target designs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗