SEARCH · Search NASA

Results for “Neutron fluence”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4

All sources NASA NTRS DOE OSTI

NASA NTRS2011

Optical Sensors for Monitoring Gamma and Neutron Radiation

For safety and efficiency, nuclear reactors must be carefully monitored to provide feedback that enables the fission rate to be held at a constant target level via adjustments in the position of neutron-absorbing rods and moderating coolant flow rates. For automated reactor control, the monitoring system should provide calibrated analog or digital output. The sensors must survive and produce reliable output with minimal drift for at least one to two years, for replacement only during refueling. Small sensor size is preferred to enable more sensors to be placed in the core for more detailed characterization of the local fission rate and fuel consumption, since local deviations from the norm tend to amplify themselves. Currently, reactors are monitored by local power range meters (LPRMs) based on the neutron flux or gamma thermometers based on the gamma flux. LPRMs tend to be bulky, while gamma thermometers are subject to unwanted drift. Both electronic reactor sensors are plagued by electrical noise induced by ionizing radiation near the reactor core. A fiber optic sensor system was developed that is capable of tracking thermal neutron fluence and gamma flux in order to monitor nuclear reactor fission rates. The system provides near-real-time feedback from small- profile probes that are not sensitive to electromagnetic noise. The key novel feature is the practical design of fiber optic radiation sensors. The use of an actinoid element to monitor neutron flux in fiber optic EFPI (extrinsic Fabry-Perot interferometric) sensors is a new use of material. The materials and structure used in the sensor construction can be adjusted to result in a sensor that is sensitive to just thermal, gamma, or neutron stimulus, or any combination of the three. The tested design showed low sensitivity to thermal and gamma stimuli and high sensitivity to neutrons, with a fast response time.

Boyd, Clark D.↗

DOE OSTI2024

Advanced Test Reactor Neutron Dosimetry Report

Results of neutron dosimetry measurements in ATR for cycle 173A-1 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

173A↗

NASA NTRS1972

Rare-gas analyses on neutron irradiated Apollo 12 samples.

Argon, krypton, and xenon from stepwise heating of five Apollo 12 rocks that had been irradiated to a neutron fluence of about 10 to the 19th neutrons/sq cm were analyzed mass-spectrometrically. The Ar40-Ar39 ages were determined and range from 3.18 plus or minus 0.06 x 10 to the 9th years to 3.32 plus or minus 0.06 x 10 to the 9th years. Trace elements Ba, Br, I, and U were measured via the neutron-induced reactions that produce isotopes of krypton and xenon. The following concentration ranges were observed: Ba, 46 to 70 ppm; Br, 48 to 146 ppb; I, 16 to 73 ppb; and U, 170 to 247 ppb.

Alexander, E. C., Jr.↗

DOE OSTI2024

An approach to combine neutron and ion irradiation data to accelerate material qualification for nuclear reactors

Next-generation nuclear power plants are generally characterized by higher operating temperatures, increased neutron fluences and energies, and distinct corrosive coolant environments versus the existing light water reactor fleet. Whether using existing materials in new environments, newly developed materials tailored for these environments, or new manufacturing methods, the traditional decades-long approach for materials qualification does not facilitate rapid deployment. Ion irradiation has demonstrated success in reproducing material microstructure and select property evolution resulting from neutron irradiation with three to four orders of magnitude reduction in time and cost, making it an ideal candidate for accelerated irradiation testing. Because microstructure has a large impact on bulk material properties, limited neutron irradiation data at lower damage levels can in principle be combined with accelerated ion testing results and modeling and simulation to form an accurate prediction of microstructure evolution and select properties under different neutron irradiation conditions and at higher damage levels. The objective of this work is to present a conceptual framework of specific steps to fulfill several technical challenges associated with qualifying materials for performance in radiation environments on an accelerated time frame. A brief review of the regulatory landscape for materials in nuclear environments is presented, followed by additional overviews to understand the current state of the art for correlation of materials properties across radiation environments using experimental and computational methodologies. Finally, the roles of academia, national laboratories, and industry in the advancement of this accelerated materials qualification framework are discussed as a path forward, with possible case studies presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Insulation Activation and Contamination in Low-Power Experiments

Among the diverse advanced nuclear reactor concepts being developed are designs that expect to operate at much higher temperatures than conventional light-water reactors, and therefore require alternative materials for components capable of withstanding these extreme temperatures. Insulation is one such component where alternative high-temperature industrial composites are being considered. Low-power experiments can provide an opportunity for advancing our understanding of material behavior under irradiation; however, these experiments are not without occupational hazards. The focus of this paper is the potential for activation of insulating composites when exposed to neutron fluence during reactor experiments, and to discuss the subsequent contamination of the testing area. This paper will also discuss practical measures for preventing the creation and inhalation of activated particulate matter (i.e., dust), such as engineering and administrative controls and personal protective equipment. This paper evaluates the suitability of several insulation materials for use in reactor experiments based on their activation from neutron irradiation. The materials considered are Pyrogel XTE by Aspen Aerogel, Cerablanket and Kaowool by Morgan Advanced Materials, Thermo-12 Gold by Johns Manville, and Maxsil CF6-2000 by McAllister Mills. Although these materials are not yet widely used in the nuclear industry, they are common in other industries due to their thermal insulating properties. These materials produce dust while handling that can pose health and safety hazards, which could be further exacerbated by activation. The analysis presented in this paper utilizes a deterministic method developed to predict the activation source term of the provided insulation materials. The method calculates the atomic density of each element within the composites and simulates their exposure to a specified neutron flux density representative of low power experimental conditions. During the irradiation phase and throughout the extensive cooling period, neutron absorptions and nuclear decays are tracked for relevant isotopes. The residual activity within each composite is then assessed to identify the material with the lowest potential radiological hazard to workers. Dust will be created during installation and removal of insulating materials and may also be generated during the experiment. Contamination of surrounding surfaces due to dispersion is highly likely unless preventative measures are taken. The results of this analysis are used to estimate an internal dose from inhalation of activated dust, which is significantly lower than the allowable whole-body dose. However, the risk of internal exposure may increase if the dust becomes a transport vector for nuclear material. The goals of this paper are to enhance the understanding of the potential activation and contamination hazards from the use of these insulating materials in low-power nuclear experiments and to foster safer work environments through informed decision-making and strategic planning. This work ultimately contributes to a broader understanding of how new non-nuclear materials for advanced reactors impact the radiological safety of co-located workers, helping to ensure the responsible management of radioactive materials during installation and decommissioning activities.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

NASA NTRS1979

Experimental setup for decomposition of UF6

The rate at which UF6 decomposes into UF5 and UF4 was determined as a function of neutron fluence. To study UF6 decomposition rate, an absorption cell for VUV and the associated VUV spectroscopy system was used to measure UF6 while under irradiation. The cell contains 50 torr of UF6 at room temperature to insure that the UF6 is in a gaseous state. It is shown that by determining the absorption coefficient below 2100A for UF6, above 2100A for F2, and at 4100A for cell degradation, UF6 decomposition can be measured at a neutron flux of 10 to the 12th power/sq cm for 72 continuous hours.

Rowe, M.↗

NASA NTRS2001

Neutron environments on the Martian surface

Radiation is a primary concern in the planning of a manned mission to Mars. Recent studies using NASA Langley Research Center's HZETRN space radiation transport code show that the low energy neutron fluence on the Martian surface is larger than previously expected. The upper atmosphere of Mars is exposed to a background radiation field made up of a large number of protons during a solar particle event and mixture of light and heavy ions caused by galactic cosmic rays at other times. In either case, these charged ions interact with the carbon and oxygen atoms of the Martian atmosphere through ionization and nuclear collisions producing secondary ions and neutrons which then interact with the atmospheric atoms in a similar manner. In the past, only these downward moving particles have been counted in evaluating the neutron energy spectrum on the surface. Recent enhancements in the HZETRN code allow for the additional evaluation of those neutrons created within the Martian regolith through the same types of nuclear reactions, which rise to the surface. New calculations using this improved HZETRN code show that these upward moving neutrons contribute significantly to the overall neutron spectrum for energies less than 10 MeV.

NASA Center LaRC↗

NASA NTRS1983

Neutron-transmutation-doped germanium bolometers

Six slices of ultra-pure germanium were irradiated with thermal neutron fluences between 7.5 x 10 to the 16th and 1.88 x 10 to the 18th per sq cm. After thermal annealing the resistivity was measured down to low temperatures (less than 4.2 K) and found to follow the relationship rho = rho sub 0 exp(Delta/T) in the hopping conduction regime. Also, several junction FETs were tested for noise performance at room temperature and in an insulating housing in a 4.2 K cryostat. These FETs will be used as first stage amplifiers for neutron-transmutation-doped germanium bolometers.

Palaio, N. P.↗

DOE OSTI2024

Response of 11 B enriched ZrB 2 ultra-high temperature ceramic to neutron irradiation at elevated temperatures

ZrB 2 , an ultra-high temperature ceramic (UHTC) is being considered for use in fusion reactor first-wall structures, yet its response to irradiation remains poorly understood. This study employed scanning/transmission electron microscopy (S/TEM), synchrotron X-ray diffraction (XRD), finite element calculations, and thermal property measurements to thoroughly investigate the neutron-irradiation effects on 11 B-enriched ZrB 2 . Neutron irradiations were conducted at 220 °C and 620 °C, with a neutron fluence of 2.2 × 10 25 neutron/m 2 (energy > 0.1 MeV), resulting in 3.9 dpa and 4200 appm He. The study revealed the unusual prevalence of prism loops and a > c anisotropic lattice swelling, likely linked to the low c/a ratio of ZrB 2 , leading to grain boundary microcracking. Reducing the grain sizes was effective in reducing intergranular cracking and macroscopic swelling. The observation of cavities in ZrB 2 irradiated at 620 °C, as opposed to 220 °C, prompts questions about the temperature at which vacancies in ZrB 2 become mobile, and the role of neutron absorption by 10 B in elevating irradiation temperatures. Isotopic enrichment in 11 B proves to be a viable strategy for mitigating helium production in transition-metal diborides, which is a critical consideration for nuclear applications. Irradiation-induced defects reduce the thermal diffusivity and conductivity of ZrB 2 by a factor of 4–9, which has important implications for its role as a plasma-facing material in fusion reactors that drive high heat fluxes through first-wall materials. Here, this comprehensive study lays the foundation for understanding ZrB 2 behavior under neutron irradiation and highlights important phenomena to consider for various material applications.

36 MATERIALS SCIENCE↗

NASA NTRS1978

Power density in direct nuclear-pumped He-3 lasers

The interaction of neutron beams with He-3 gas is of interest for nuclear pumped lasers. The effects of spectral dependence of the neutron beam, neutron attenuation in the gas-filled laser tube, and transport of the charged-particle He-3(n, p)He-3 reaction products are treated in detail. An expression for the energy density as a function of position within the tube, tube radius, operating pressure, and neutron fluence is given. The maximum energy density within the optical cavity is achieved when the tube radius is 3.26/P where P (atm) is the operating pressure. The variation of radius by 50% above and below optimum will change the energy density at most by 10%, although performance degrades quickly for radii outside this range. If the optimum tube radius is used for each operating pressure, then the power density on the centerline (kW/cu cm) is given as 9.3 x 10 to the -18th power times the operating pressure times flux in a thermal neutron environment of flux (n/sq cm sec).

Wilson, J. W.↗

NASA NTRS1996

In-flight radiation measurements on STS-60

A joint investigation between the United States and Russia to study the radiation environment inside the Space Shuttle flight STS-60 was carried out as part of the Shuttle-Mir Science Program (Phase 1). This is the first direct comparison of a number of different dosimetric measurement techniques between the two countries. STS-60 was launched on 3 February 1994 in a nearly circular 57 degrees x 353 km orbit with five U.S. astronauts and one Russian cosmonaut for 8.3 days. A variety of instruments provided crew radiation exposure, absorbed doses at fixed locations, neutron fluence and dose equivalent, linear energy transfer (LET) spectra of trapped and galactic cosmic radiation, and energy spectra and angular distribution of trapped protons. In general, there is good agreement between the U.S. and Russian measurements. The AP8 Min trapped proton model predicts an average of 1.8 times the measured absorbed dose. The average quality factor determined from measured lineal energy, y, spectra using a tissue equivalent proportional counter (TEPC), is in good agreement with that derived from the high temperature peak in the 6LiF thermoluminescent detectors (TLDs). The radiation exposure in the mid-deck locker from neutrons below 1 MeV was 2.53 +/- 1.33 microSv/day. The absorbed dose rates measured using a tissue equivalent proportional counter, were 171.1 +/- 0.4 and 127.4 +/- 0.4 microGy/day for trapped particles and galactic cosmic rays, respectively. The combined dose rate of 298.5 +/- 0.82 microGy/day is about a factor of 1.4 higher than that measured using TLDs. The westward longitude drift of the South Atlantic Anomaly (SAA) is estimated to be 0.22 +/- 0.02 degrees/y. We evaluated the effects of spacecraft attitudes on TEPC dose rates due to the highly anisotropic low-earth orbit proton environment. Changes in spacecraft attitude resulted in dose-rate variations by factors of up to 2 at the location of the TEPC.

NASA Discipline Radiation Health↗

DOE OSTI2026

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175C: Results of Neutron Dosimetry Measurements for the Advanced Test Reactor Cycle 175C

Results of neutron dosimetry measurements in ATR for cycle 175C are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗

DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report - 173B

Results of neutron dosimetry measurements in ATR for cycle 173B-1&2 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

173B↗

DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175A-1

Results of neutron dosimetry measurements in ATR for cycle 175A-1 are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

175A↗

DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report - 173B (Rev. 1)

173B↗

DOE OSTI2026

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175B-1

Results of neutron dosimetry measurements in ATR for cycle 175B-1 are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

175B↗

DOE OSTI2025

Neutron capture signatures determined from gadolinium isotope compositions of uranium ore concentrates using MC-ICPMS

We present Gd isotope compositions on 25 well-characterized uranium ore concentrates (UOCs). About half the UOCs have isotope depletions in 157 Gd coupled with anticorrelated excesses in 158 Gd due to neutron capture effects. UOCs from older ore bodies and with higher U contents show larger neutron capture effects than UOCs from younger ore bodies with lower U contents. These Gd data are correlated with their previously measured Sm isotope compositions, and we use the Sm data to estimate the neutron fluence of these samples. In conclusion, this work demonstrates how Gd isotope signatures can be employed as a new tool for nuclear forensics.

Earth Sciences↗