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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

STAYSL PNNL Suite

The STAYSL PNNL Suite of software provides a set of tools for working with neutron activation rates measured in a nuclear fission reactor, an accelerator-based neutron source, or any neutron field to determine the neutron flux spectrum through a generalized least-squares approach. This process is referred to as neutron spectral adjustment since the preferred approach is to use measured data to adjust neutron spectra provided by neutron physics calculations. The input data consist of the reaction rates based on measured activities, an initial estimate of the neutron flux spectrum, neutron activation cross sections and their associated uncertainties (covariances), and relevant correction factors. The output consists of the adjusted neutron flux spectrum and associated covariance matrix, which is useful for neutron dosimetry and radiation damage calculations. The software suite consists of the STAYSL PNNL, SHIELD, BCF, and NJpp Fortran codes and the SigPhi Calculator spreadsheet tool. In addition, the development of this software suite and associated data libraries used the third-party NJOY99 Fortran code (http://t2.lanl.gov/nis/codes/njoy99/). The NJOY99 and NJpp codes are used to assemble cross section and covariance input data libraries (for both SHIELD and STAYSL PNNL) from the International Reactor Dosimetry File of 2002 (IRDF-2002; http://www-nds.iaea.org/irdf2002/) developed by the Nuclear Data Section of the International Atomic Energy Agency (Vienna, Austria). The BCF, SigPhi Calculator, and SHIELD software tools are used to calculate corrected activation rates and neutron self-shielding correction factors, which are inputs to the STAYSL PNNL code.

Greenwood, L.R.↗
DOE OSTI2018

Enhancement of STAYSL_PNNL with IRDFF/V1.05 to 60 MeV

The STAYSL PNNL program performs a least-squares adjustment of the neutron spectrum based on the IRDFF cross section library and user-defined activation data and a starting neutron spectrum. All known covariances are included. Required corrections are included for the irradiation history, gamma absorption, neutron burn-up, and neutron self-shielding using separate utility programs prior to the execution of STAYSL PNNL. Several new neutron cross section libraries based on IRDFF V1.05 to 60 MeV are provided with the program including energy group structures of 69 (WIMS), 129 (14 MeV), 140 (prior 100 groups extended to 60 MeV), 175 (Vitamin J), 640 (old SANDII to 20 MeV), and 725 (old Sand II extended to 60 MeV) energy groups. Twelve new reactions were included in IRDFF V1.05. At higher neutron energies, calculations demonstrate the need for activation cross sections from natural targets such as Ti(n,x)46Sc rather than only isotopic targets such as 46Ti(n,p)46Sc and 47Ti(n,x)46Sc. The effects of such reactions increase dramatically with increasing neutron energies and cannot be neglected. Some critical reactions such as 56Fe(n,x)54Mn are not currently included in IRDFF and prevent the use of the 54Fe(n,p)54Mn reaction at higher neutron energies. It is recommended that natural element cross sections be included in future releases of IRDFF for a list of reactions that suffer from reactions on multiple isotopes that lead to the same activation product. Spectral adjustments are illustrated in Be(d,n) thick target neutron fields at deuteron energies of 30 and 40 MeV.

reactor dosimetry, neutron spectral adjustment↗
DOE OSTI2018

Retrospective neutron spectrum determination of a (30 MeV D, Be) source using the multi-foil activation technique and STAYSL-PNNL [Characterization of neutron spectra from a (30 MeV d, Be) source using the multi-foil activation technique]

Retrospective characterization of a (30 MeV D, Be) neutron source was performed employing multi-foil activation and STAYSL-PNNL. Experimental reaction rates were calculated from gamma spectroscopy measurements of irradiated foils and MCNP provided the guess spectrum. Adjusted spectra were evaluated through activation calculations for a stainless-steel target using FISPACT-II. Adjusted spectra showed limited dependence on the dosimetry reactions and provided minor improvements in activation calculations. Omitting reflected neutrons in the guess spectrum generated poor activation results and the limited number of dosimetry reactions introduced doubt in the adjusted spectra. Furthermore, a dedicated neutron spectrometry experiment and a more detailed simulation is required.

42 ENGINEERING↗
DOE OSTI2013

STAYSL PNNL Suite

This announcement is for the "STAYSL PNNL Suite" software developed by Pacific Northwest National Laboratory (PNNL).

Source record↗
DOE OSTI2013

User Guide for the STAYSL PNNL Suite of Software Tools

The STAYSL PNNL software suite provides a set of tools for working with neutron activation rates measured in a nuclear fission reactor, an accelerator-based neutron source, or any neutron field to determine the neutron flux spectrum through a generalized least-squares approach. This process is referred to as neutron spectral adjustment since the preferred approach is to use measured data to adjust neutron spectra provided by neutron physics calculations. The input data consist of the reaction rates based on measured activities, an initial estimate of the neutron flux spectrum, neutron activation cross sections and their associated uncertainties (covariances), and relevant correction factors. The output consists of the adjusted neutron flux spectrum and associated covariance matrix, which is useful for neutron dosimetry and radiation damage calculations.

Reactor dosimetry, neutron spectral adjustment↗
DOE OSTI2026

Validation and Verification of Python based Neutron Spectrum Unfolding Software

To validate and verify the python-based code (PySL), designed to replicate the programs used by STAYSL for Beam Correction Factor (BCF) and Self-Shielding Factor (SHIELD), a series of tests were performed. To test BCF a python script was written to generate a random flux history file and both versions of the code processed the data. The test verified matching values up to at least one decimal place, approximately 10,000 tests where run and each one passed. Isotopes began to fail the tests once neutron saturation was reached. To verify this the total time of exposure was varied the isotopes that failed were compared to a list of their half-lives. The test process for SHIELD was very similar but, in this case, the code began by producing an input file with varying thickness and device type/environment for the SHIELD input. The failure condition for this test was if any of the data points for an isotope had a difference above 3%. Approximately 40 of these tests were run and there were only 3 isotopes that had reoccurring failures but only 2% of their points were above the 3% difference. A visual comparison was conducted by plotting the results from both programs. Although the test failed, the differences between their values were minuscule, and the self-shielding factor’s shape was preserved when plotted. Next steps for this project will be validating and verifying the python-based SigPhi code and then reproducing and testing the least squares unfolding performed by STAYSL.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Uranium integral fission product yields for a spectrally-shaped 14.1 MeV neutron source at the National Ignition Facility

This paper describes the experimental results for an energy tuning assembly created to modify the National Ignition Facility deuterium–tritium fusion neutron source into a notional thermonuclear and prompt fission neutron spectrum, which has applications in integral measurements, nuclear data benchmarks, and radiation effects on microelectronics. The Monte Carlo neutron transport utilized MCNP5 to estimate the ETA-modified fluence using the ENDF-B/VIII.0 and IRDFF-II continuous energy nuclear data libraries, and SCALE Sampler was used to estimate the systematic nuclear data covariance using ENDF-B/VII.1 and IRDFF-II in a 252-group structure. The experiment fielded eight activation foils and a highly enriched uranium sample. This provided fifteen reaction channels that were used in a forward-fit comparison to the modeled results and to unfold the neutron spectrum using STAYSL. Gamma-ray spectrometry was performed on the activation and highly enriched uranium foils, and the reduced Χ 2 between the modeled and experimental values was 1.21. The results from the STAYSL unfold, reduced Χ 2 = 1.62, indicated that the modeled neutron spectrum was achieved and the systematic nuclear data uncertainty associated with the neutron transport and activation product cross sections was representative of the experiment. Integral cumulative fission product yield data were collected for 37 mass chains with a combination of gamma-ray spectrometry and radiochemical analysis. Fission product analysis was generally in agreement with two models using a semi-empirical fit and the General Observables of Fission code, with the exception of mass chains 88, 109, 111, 112, 113, 129, 139, 142, 144, 151, and 156.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Physics-informed Estimation of the Covariance Matrix for Various Neutron Spectra

A method for estimating covariance matrices which capture the uncertainties in calculated reactor spectra has been developed. This method is based on perturbing the parameters of a physics-based analytic model fitted to a calculated spectrum. The covariance of the perturbed analytic spectra imposes energy-dependent correlations due to the physics of the neutron processes in the reactor, i.e., a fission component, a 1/E down-scatting component, and a thermal Maxwellian component. An analytic model is developed which is shown to produce good fits to several reactor environments. The covariance matrices produced via this method are then used as the prior spectrum in STAYSL least squares spectrum adjustment where it is combined with integral metrics, such as activation measurements, to produce a high-fidelity neutron spectrum characterization. It was concluded that the methodology showed agreeable results for the ACRR free-field spectrum adjustment in STAYSL resulting in a 𝜒 2 value of 2.21 (per degree of freedom), but further work is needed to describe scattering and interface regions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Characterization of the Fast-Neutron Irradiator and the Fast-Flux Tube Irradiation Fixtures at the Pennsylvania State Breazeale Reactor

Accurate knowledge of the neutron spectrum at a nuclear research reactor is a prerequisite for planning irradiation experiments, as well as for evaluating irradiation exposure results. The neutron-flux spectrum in the fast-neutron irradiator (FNI) and the fast-flux tube (FFT) irradiation fixtures at the Pennsylvania State Breazeale Reactor (PSBR) were characterized using the multi-foil neutron activation method. These irradiation fixtures make use of graded shielding to produce unique neutron fields. Multiple foil sets were irradiated in the fixtures with different exposure times and reactor powers to understand the stability over a wide range of operating conditions. Measured results were evaluated against a MCNP6 simulation to produce a measurement-informed neutron flux-energy spectrum for each fixture using STAYSL_PNNL. Simulated estimates of the FNI fixture, a newer fixture (~25 years old), demonstrated excellent agreement with measured results; the FFT did not. Thermal neutron measurements from the FFT suggest there is additional thermal leakage not captured in the simulation model. In conclusion, possible explanations for the discrepancy include burn-out or degradation (i.e., micro-cracking, gaps, etc.) in the boral and cadmium liners over the lifetime of the fixture (~40 years old).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Characterization of the neutron flux during production of 18 F at a medical cyclotron and evaluation of the incidental neutron spectrum for neutron damage studies

The fast neutron spectrum was measured an irradiation position adjacent to two 18 F (fluorine-18) production targets at a medical cyclotron. The neutron spectrum was modeled using MCNP6 and experimentally determined using activation foils. A three-group neutron spectrum was determined using an over-determined least squares fitting method. The MCNP6 model results and measurements of the activation foils were used as input to unfold the neutron spectrum using the PNNL-STAYSL code. The neutron spectrum adjacent to the 18 F production target was evaluated for use in incidental experiments to evaluate neutron induced radiolysis of polyurethane and radiation hardness testing of electronics. The fast neutron flux near the 18 F target between the energies of 0.4 – 16.5 MeV was 1.8 – 3.0x10 9 n cm -2 s -1 . The neutron dose rate in polyurethane at the irradiation position was 2.8 cGy s -1 – 5.8 cGys -1 . The 1 MeV equivalent neutron flux was 2.4x10 9 n cm -2 s -1 – 4.0x10 9 n cm -2 s -1 .

18F-fluorodeoxyglucose↗
DOE OSTI2019

Performance evaluation of an energy tuning assembly for neutron spectral shaping

An energy tuning assembly (ETA) was designed here to be fielded at the National Ignition Facility (NIF) to modify the characteristic D-T fusion spectrum to include a prompt fission neutron spectral component. The ETA was characterized at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory to measure the shaped spectrum from an incident deuteron breakup neutron source, test the proposed neutron spectroscopy techniques used to inform the flux measurements at NIF, and validate the ability to predict ETA performance using a Monte Carlo Neutral Particle (MCNP) simulation. Activation foils (i.e., Ni, In, Au, Al) were exposed to a collimated 33-MeV deuteron-breakup beam originating from a tantalum breakup target. The source spectrum absent the ETA was characterized using a set of activation foils and the STAYSL unfolding code. Finally, the ETA-modified spectrum was obtained using activation foil unfolding with a χ 2 ν = 0 . 92 . The ETA-modified unfolded spectrum agreed with the MCNP-simulated prediction in the energy range of 0.1–14 MeV, but exhibited disagreements in the 10 eV–100 keV region. This work demonstrates shaping of the NIF neutron spectrum via the ETA to be a viable path forward for tailored neutron beams at NIF.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Determining the 9 $\mathrm{Be}$($n, γ$) 10 $\mathrm{Be}$ integral cross section at fission neutron energies

The 9 B neutron capture cross section has significant implications for Be materials in the nuclear industry as well as the α process in stellar nucleosynthesis. While the cross section is well constrained at thermal neutron energies, there is a lack of experimental data at higher neutron energies, and the evaluated nuclear data libraries can differ by up to two orders of magnitude. We calculate the 9 Be(n, γ) 10 Be integral cross section at fission neutron energies in an effort to resolve disagreements amongst the nuclear data libraries. Foil irradiation experiments were performed using the Flattop critical assembly at the National Criticality Experiments Research Center with either the highly enriched U or Pu cores, with target foil stacks placed at multiple locations to exploit different neutron energy profiles. Accelerator mass spectrometry was used to measure the 10 Be/ 9 Be ratio in irradiated Be foils, while all other activation products were quantified through gamma spectrometry. The experiments were simulated using the Monte Carlo N-Particle radiation transport code and combined with experimental results to determine the total neutron fluence, while the staysl-pnnl suite and fispact-ii code were used to validate the model and assess the systematic uncertainty. The new 9 Be(n, γ) 10 Be integral cross sections calculated in this work are 26.5 ± 2.2µb at 0.59 ± 0.07 MeV, 24 ± 3 µb at 0.98 ± 0.14 MeV, 21.7 ± 1.3 µb at 1.26 ± 0.11 MeV, 21.8 ± 1.4 µb at 1.32 ± 0.11 MeV, and 18.6 ± 1.1 µb at 1.46 ± 0.13 MeV. These results do not agree with integral cross sections from any of the nuclear data library evaluations. Discrepancies between the new integral cross sections reported here and the nuclear data libraries suggest a more complex cross-section structure in the MeV range which allows for more resonance contributions, and more work is needed to further constrain the evaluated cross sections.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Characterization of D-T Generator Neutron Flux Spectrum for Cyclic Neutron Activation Analysis Experiments

Improving nuclear data for short-lived fission product yields will further our fundamental understanding of fission, which is needed across various scientific fields and applications. One method of attaining the needed product yield data is through cyclic neutron activation, which allows a target to be irradiated in a neutron environment and then transported for counting of the radionuclides produced, typically via g spectroscopy. Recently, such a system has been constructed and commissioned at Pacific Northwest National Laboratory. Targets are shuttled between the head of a D-T neutron generator and a counting station with a transit time of 2 seconds. As part of the characterization of this system, the neutron flux was studied using two activation targets. The neutron flux from the deuterium-tritium fusion generator was determined to be 9.95(33) x10^8 n/cm2-s with a peak energy of 14.9 MeV and a spread of approximately 4 decades between the epithermal and 14 MeV peak group flux.

cyclic neutron activation, gamma-ray spectroscopy,↗
DOE OSTI2023

Godiva IV Central Cavity Neutron Environment Characterization with Threshold Neutron Detectors

Godiva IV is a cylindrical fast burst reactor comprised of approximately 65 kg of highly enriched uranium that is operated by Los Alamos National Laboratory and sited at the National Criticality Experiments Research Center at the Nevada National Security Site in Nevada in the United States. Godiva IV is typically operated at delayed critical and in the regime spanning from sub-prompt to super-prompt bursts. Godiva IV is used for sample irradiations, criticality safety demonstrations, dosimetry studies, and for studying super-prompt behavior. In preparation for both an upcoming experiment to reduce uncertainties in the prompt fission spectrum for 235 U using threshold neutron detectors, and for future research using Godiva IV, it was desired to exercise the process of the selection of threshold neutron detectors/activation foils, radiation metrology, and the subsequent adjustment of the neutron spectrum. For this exercise, nine high purity threshold neutron detectors/activation foils were irradiated in a Godiva IV burst. The foils were then analyzed using a high purity germanium detector in the NCERC counting laboratory to determine end of irradiation specific activities for available IRDFF-II reactions. This work summarizes the Godiva IV foil irradiation, radiation metrology results, and adjusted neutron spectrum. The results of this exercise ultimately characterized the neutron environment inside the sample irradiation cavity inside Godiva IV to a higher degree than previously performed, informed decisions for the upcoming larger scale experiment, and will inform future neutron spectrum characterizations at NCERC.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗