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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

In-Situ Gamma-Ray Assay of the 235-F Plutonium Fuel Form Facility at the Savannah River Site - 16293

The Plutonium Fuel Form (PuFF) Facility is located in Building 235-F near the geographic center of the Savannah River Site. Between 1978 and 1984, the facility was used to produce approximately 165 kilograms of iridium-encapsulated Pu-238 spheres and pellets for use as radioisotope thermal generators, primarily for the space program. The facility is divided between two cell lines, the east cell line used to process powdered Pu-238 oxide raw material into fuel forms and the west cell line used to encapsulate the fuel forms in iridium. Scoping in-situ gamma-ray assays were performed in the PuFF Facility in 2006 and estimated that hundreds of grams of Pu-238 holdup were within the facility. Using this holdup estimate as a source term, SRS has performed a risk analysis that indicated a seismic event that induces a full-facility fire in 235-F could lead to a 28,800 rem dose to a co-located worker. Based on this risk assessment, SRS is taking steps to decontaminate the facility. One of the first steps taken has been to improve upon the quality of the in-situ gamma-ray assay data. Carts and collimators were specially designed to survey the equipment in the PuFF Facility. The cell interiors were assayed along with the furnaces and storage coolers that protrude beneath the cells. While the previous scoping work consisted of 32 measurements, the current series of measurements included nearly 400 measurements using both high-purity germanium and lanthanum bromide detectors. Data analysis for the current set of measurements was conducted with greater rigor as well. MCNP5 was used to evaluate a variety of possible physical distributions for the Pu-238 source term and estimate cross-talk between neighboring cells. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. This also allowed the application of matrix correction factors to adjust gamma-ray attenuation. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the PuFF facility. This work lowered the estimate of holdup in the PuFF facility by approximately 100 grams compared to the previous estimate, a 0.55's difference. The uncertainty in the Pu-238 holdup was also reduced substantially relative to the 2006 scoping assay. Further radiological measurements are planned for the near future including in-situ gamma-ray assays of wing cabinets in the east cell line maintenance area, the transfer tube between the two cell lines, and some of the furnaces and storage coolers. A germanium gamma imager, the GeGI, will also be used to map out the distribution of the Pu-238 source term in the PuFF Facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

2015 In-Situ Gamma-Ray Assay of the West Cell Line in the 235-F Plutonium Fuel Form Facility

In November and December 2015, scientists from SRNL took a series of in-situ gamma-ray measurements through the windows in front of Cells 6-9 on the west line of the PuFF facility using a shielded, 120% high-purity germanium detector. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Where possible, the distribution of the Pu-238 in the cells was determined using the Germanium Gamma-ray Imager (GeGI). This distribution was then fed into the MCNP model to quantify the Pu-238 in each cell. Data analysis was performed using three gamma rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

In-Situ Gamma-Ray Assay of the West Cell Line in the 235-F Plutonium Fuel Form Facility

On August 29 th , 2013, scientists from SRNL took a series of in-situ gamma-ray measurements in the maintenance trench beneath Cells 6-9 on the west line of the PuFF facility using an uncollimated, high-purity germanium detector. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the West Cell Line of PuFF. The results of the assay measurements are found in the table below along with the results from the scoping assay performed in 2006. All uncertainties in this table (as well as the rest of the report) are given as 1σ. The total holdup in the West Cell Line was 2.4 ± 0.7 grams. This result is 0.6 g higher than the previous estimate, a 0.4σ difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

In-Situ Gamma-Ray Assay of the East Cell Line in the 235-F Plutonium Fuel Form Facility

On September 17 th -19 th , 2013, scientists from SRNL took a series of in-situ gamma-ray measurements in the maintenance trench beneath Cells 1-5 on the east line of the PuFF facility using a well-collimated, high-purity germanium detector. The cell interiors were assayed along with the furnaces and storage coolers that protrude beneath the cells. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the East Cell Line of PuFF. The results of the assay measurements are found in the table on the following page along with the results from the scoping assay performed in 2006. All uncertainties in this table (as well as the rest of the report) are reported at 1σ. Summing the assay results and treating MDAs as M 238Pu = 0 ± MDA, the total holdup in the East Cell Line was 240 ± 40 grams. This result is 100 grams lower than the previous estimate, a 0.55σ difference. The uncertainty in the Pu-238 holdup is also reduced substantially relative to the 2006 scoping assay. However, the current assay results are in agreement with the 2006 scoping assay results due to the large uncertainty associated with the 2006 scoping assays. The current assay results support the conclusion that the 2006 results bound the Pu-238 mass in Cells 1-5. These results should be considered preliminary since additional measurements of the East Cell line are scheduled for 2017 and 2018. Those measurements will provide detailed information about the distribution of Pu-238 in the cells to be used to refine the results of the current assay.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Gamma-ray imaging assay of cells 3-5 of the east cell line in the 235-F plutonium fuel form facility

In August and September, 2016, scientists from the Savannah River National Laboratory (SRNL) took a series of gamma-ray imaging measurements through the cell windows in front of Cells 3-5 on the east line of the Plutonium Fuel Form (PuFF) Facility using an electrically cooled, high-purity germanium detector. A Germanium Gamma Ray Imager (GeGI) was utilized since it allowed for the location from which the radiation was being emitted to be identified by incoming gamma-ray energy. This measurement technique provided a tool which allowed for the relative concentration of Pu-238 to be mapped throughout each cell. The mapping and new assay data were then used to update the model used in an assay discussed in a 2014 report (SRNL-STI-2014-00629) and to calculate a more accurate value for the holdup in each of the cells [1]. Note that the mapping and new assay data did not replace the previous assay data in the model. Rather, the mapping and new assay data provided additional details on source distribution, which supplemented the previous assay data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Selected Profiles of Burnup and Isotopic Concentrations in Pebble Bed High Temperature Reactor Fuel

Pebble Bed High Temperature Reactors (PB-HTRs) feature a unique fuel loading, burnup, and discharge process that is markedly different that a light water reactor or any other batch-loaded reactor that uses fixed fuel elements. During PB-HTR operation, small spherical fuel elements are dropped onto the top of the bed at small and regular intervals, 'trickle' through the core, and pass out of the core via a discharge chute at the bottom. Partially burnt pebbles are circulated back to the top of the core for another pass through the flux field. The process continues until the pebble achieves a specified discharge burnup. This results in high fuel utilization relative to a comparable batch-loaded fuel assembly which may or may not achieve full burnup depending upon where it is placed in the core. On the other hand, the radial and axial variation in the flux field and the random placement of each element on top of the bed mean that there will be a distribution of the final burnup and isotopic concentrations in spent fuel. This leads to some uncertainty in burnup analysis that has consequences for burnup monitoring (and thus efficient operation), source term quantification, and material accountability. The PEBBED code was developed specifically for the design and analysis of PB-HTRs. Although it is still under development, the code captures the important physics of cores fueled by dropping in, and if so configured, recirculating small graphite spheres in which the Tri-Isotropic (TRISO) coated particle fuel is embedded. In this study, the capabilities of the code for computing the isotopic contents of PBR fuel were exploited to examine and compare the variability of selected isotopes, in this case U-235, Pu-239, and Cs-137, in the fuel discharged from different PBR concepts. Results indicate the sensitivity of certain isotopes to the core geometry, mainly the presence of an inner reflector. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

On the Uranium and Plutonium Nuclear Data Evaluations [Slides]

This presentation provides overviews of uraniums, 233 U and 235 U, plutonium research including motivations, current status in ENDF/B-VIII.0, the inclusion of sub-thermal data, and the inclusion of LANL ratio capture-to-fission data. Additionally, the preliminary fit of Mosby’s data as reported, fluctuating neutron multiplicities, uncertainty in evaluated libraries, and experimental effects are also presented. The presentation concludes by providing a summary of plans for the U and Pu evaluation work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Nuclear Data Sheets for A = 235

Spectroscopic data and level schemes from radioactive decay and nuclear reaction studies are presented here for all nuclei with mass number A=235. The highlight of this evaluation consists of the precise and comprehensive Coulomb excitation study (2012Wa35) on {sup 235}U, which in addition to the 7/2[743] ground state rotational band, extended the 1/2[631], 5/2[622], 5/2[752], and 3/2[631] rotational bands up to Jπ=53/2+, 49/2+, 441/2−, and 43/2+, respectively. This evaluation presents a study (2010Hu02) of the {sup 237}Np({sup 116}Sn, {sup 118}Snγ) reaction where the ground state rotational band 5/2[642] was observed up to Jπ=(53/2+). It is worth for historical knowledge to mention the report on the “Discovery of isotopes of the transuranium elements with 93≤Z≤98” (2013Fr02), where the information for elements Np, Pu, and Am with mass number A=235 is given. {sup 235}Cf has not been observed. The alpha hindrance factors (HF) presented in this evaluation were calculated using values of the radius parameter (r{sub 0}) interpolated from those for even–even adjacent nuclei given by 1998Ak04.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

A novel approach to mitigating the potential release of radioisotopes under fire conditions - enhancing fire resiliency of radiological contamination fixatives during deactivation & decommissioning activities

Savannah River National Laboratory (SRNL), in close collaboration with the Florida International University Applied Research Center (FIU ARC), successfully executed a technology development activity on behalf of the Department of Energy, Office of Environmental Management (DOE EM). The purpose of the activity was to improve the capability of fixative technologies in immobilizing residual contamination when exposed to thermal stressors as postulated in accident scenarios in Basis for Interim Operations (BIO) documents across the DOE EM complex. The effort resulted in the test and evaluation of a down-selected, commercial-off-the-shelf (COTS) intumescent technology in a radioactive environment at the Savannah River Site (SRS) Building 235-F Plutonium Fuel Form (PuFF) Facility and highlighted the potential use of this technology as a stand-alone, fire retardant fixative. Furthermore, this activity highlighted significant shortfalls in common fixatives currently used to support decommissioning activities and brought to the forefront the need for a methodical, uniformed approach to certifying fixative technologies for operational use in Deactivation and Decommissioning (D&D) activities. An essential component of this research was to evaluate the state of industry fixatives currently in use and set the foundation for comparison to alternative technological solutions. A baseline of five (5) commonly used fixatives and decontamination gels (hereafter collectively referred to as fixatives) was conducted, and notable shortfalls and deficiencies were revealed, particularly when exposed to thermal, water, and other environmental stressors. At temperatures as low as 300 °F - 400 °F (148.89 °C – 204.44 °C), all commonly used fixatives melted from the substrates within 3-5 minutes of exposure, resulting in contaminant transport. Significant degradation in terms of mass loss, desiccation, and off-gassing occurred, and the chemical breakdown of the polymer was so complete researchers assessed there would likely have been a release of residual contamination. Additional vulnerabilities became evident when exposed to water immersion and high humidity; existing fixatives took up water, swelling and frequently delaminating from the substrate, once again increasing the likelihood of a contaminant release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Stabilization of Residual Contamination with Alternative Materials - 19404

Since 2015, Savannah River National Laboratory (SRNL) in conjunction with Florida International University (FIU) have been investigating the applicability of intumescent coatings as radiological contamination fixatives for stabilization of residual materials after decontamination efforts have concluded. Intumescent coatings are designed to prevent heat transfer to a protected substrate and are typically used in both the aerospace industry to protect components from thermal shock and in building design to protect building joints that may fail in the event of a fire. For this work, these materials are examined for their ability to reduce worker risk by fixating residual contamination and lower the impact of a release in a scenario that breaches containment of a contaminated building and subsequently induces a large-scale fire by crediting their fire-retardant attributes. Initial studies of intumescent coatings and commercial fixatives included performance in direct flame, varied environmental conditions (20-110 F or -6.67 C, 10-95% relative humidity), adhesion studies to stainless steel after curing in varied environmental conditions, and radiological exposure. Of all the materials tested, many were found to be inadequate in one or more of these areas. These materials failed in one or more of three characteristic ways: 1) through delamination at reduced temperatures (≤40 F or 4.44 C) or from even moderate handling, 2) melting off the substrate at elevated temperatures (≥90 F or 32.22 C), 3) proving to be flammable and promote smoke propagation. Failure in these ways results in the potential for a subsequent release of the fixated contamination in the event of water presence/absence or direct flame. Using these results as a platform for the best in class identified material that did not experience failure in any of these tests, SRNL and FIU targeted the plutonium-238 contaminated hot cells within Savannah River Site (SRS) Building 235-F as a characteristic test bed to benchmark an intumescent coating used as a contaminant fixative. Through discussion with the SRS 235-F Risk Reduction personnel, appropriate hazards and performance metrics were discussed and a cold demonstration was performed in a hot cell mockup located at SRS. The application methods were proven functional in the cold mock up, and these intumescent coatings were subsequently utilized in the Savannah River Site Building 235-F's Plutonium Fuel Form (PuFF) Facility to aid in stabilization of plutonium-238 in both a contaminated hot cell and Entry Hood in September 2018. Presented in this work is the overview of the application process, associated working considerations for introduction of equipment and materials into the facility, lessons learned, and lifetime monitoring data to date. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

High sensitivity measurement of femtogram-level 238 Pu by thermal ionization mass spectrometry with correction of background 238 U

Low-concentration plutonium isotopic measurements provide a valuable fingerprint to identify possible material origins in the context of environmental contamination monitoring, environmental tracing, nuclear safeguards, nuclear forensics, and treaty monitoring. Thermal ionization mass spectrometry (TIMS) has long been recognized as one of the most sensitive techniques for plutonium isotopic measurement, but has not generally been utilized for 238 Pu determination, especially at low concentrations, due to an isobaric interference from 238 U, prevalent as background. Here, this work demonstrates a new analytical technique for correction of background 238 U from the 238 Pu counts collected during a TIMS analysis. The technique takes advantage of the higher ionization temperature of U relative to Pu and uses a 235 U tracer added after plutonium purification chemistry for 238 U semi-quantitative determination. This technique has been successfully applied to the determination of 238 Pu concentration and the 238 Pu/ 239 Pu ratio in sample types including single-isotope standards, isotopic certified reference materials, matrix-containing environmental reference materials, and simulated nuclear detonation debris. The results have been directly compared on a sample-by-sample basis to alpha spectrometry results. The detection limit by this new technique is 0.23 fg 238 Pu, which was previously unachievable by mass spectrometry and is possible in around 1 h of analysis time post-chemistry compared to weeks of required counting time by alpha spectrometry (at the same atom amount). We also present original, high precision data for Pu isotope concentrations in IAEA-384, Fangataufa Sediment, including 242 Pu results for the first time. The new technique allows measurement of a key isotope, 238 Pu, that was previously not measured in small environmental or nuclear forensics collections.

238Pu↗
DOE OSTI2024

$\overline{TKE}$ Parameterization and $\bar{v}$ Uncertainty Analysis for CGMF

Previous work was performed on tuning CGMF parameters for 235 U, 238 U, and Plutonium isotopes. Now work is being done to tune minor uranium isotopes. However, uranium isotopes like 232 U and 236 U have almost no experimental data. We are applying cross-isotope models to extrapolate and tune CGMF on isotopes that lack experimental data. There exist several internal CGMF physics quantities that affect the output of CGMF—multi-chance fission probability, excitation energy sharing, spin-cutoff factor, spin scaling, and fragment total kinetic energy to name a few. The mean fragment total kinetic energy, $\overline{TKE}$, is particularly interesting because of its strong anti-correlation with $\bar{v}$. We are most interested in the mean fragment total kinetic energy before neutron emissions. $\overline{TKE}$ is assumed to be pre-neutron emission unless otherwise stated. Currently in CGMF, the $\overline{TKE}$ model for 233,234,235,238 U are tuned independently to reproduce ν for the associated isotopes. In this report, we will tune a cross-isotope $\overline{TKE}$ model to experimental $\overline{TKE}$ data for 232,233,234,235,236,238 U. Because of the unreliable and sparse nature of $\overline{TKE}$ experimental data, future work will use more reliable experimental $\bar{v}$ data to infer the $\overline{TKE}$ model (and likely other internal CGMF parameters) for uranium isotopes. Such work has been performed previously using a sensitivity analysis and Kalman filter methods.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Quasi-differential neutron induced neutron emission reaction measurements at WNR

The quasi-differential neutron induced neutron emission reaction measurement (QD-NINE) method provides an additional benchmark which constrains cross-section evaluations and provides a measure of verification for existing evaluations. Experiments are preformed using an array of EJ-309 liquid scintillators at a variety of angles measuring the neutron scattering response from a sample.. An MCNP simulation is then compared to this experimental data and is run using various sets of evaluated nuclear data. Since the isotopes of interest for this measurement, {sup 235}U and {sup 239}Pu, are difficult to obtain, the WNR facility at LANL was chosen to perform the measurements utilizing the chi-nu detector array. A measurement of carbon has been performed and the high energy cutoff has been extended from 20 to 60 MeV compared with previous measurements using this method. An early stage simulation was compared to the experimental data and the results show good agreement. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Chi-Nu measurements of the prompt fission neutron spectrum of {sup 235}U and {sup 239}Pu - 25581

Knowledge of the prompt fission neutron spectrum is crucial for understanding the behavior of energetic nuclear systems. Although this energy spectrum has been measured for the major actinides {sup 235}U and {sup 239}Pu for decades, the majority of the available literature data lack the coverage of both incident and outgoing neutron energies and the precision required to accurately guide evaluations and fission models. The Chi-Nu experiment is designed to measure the prompt fission neutron spectrum for a wide range of incident and outgoing neutron energies with attention to uncertainty estimation and assessment of systematic errors in order to provide reliable guidance for evaluations and models of nuclear fission. Preliminary results of the prompt fission neutron spectrum from the neutron-induced fission of {sup 235}U and {sup 239}Pu are presented here, with a focus on the presence of spectral features near multi-chance fission thresholds, pre-equilibrium neutrons in the deduced spectrum, and also on trends of the mean fission neutron energy as a function of the incident neutron energy. While the mean energies of from ENDF/B-VIII.0 are in good agreement with those reported for {sup 235}U spectra, significant disagreements exist above an incident neutron energy of approximately 13 MeV for {sup 239}Pu. Furthermore, in general, the physics features regarding multi-chance fission and pre-equilibrium neutron emission newly-introduced into ENDF/B-VIII.0 appear provide an improved description of the data as compared to ENDF/B-VII.1. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2011

A Second Look at Neutron Resonance Transmission Analysis as a Spent Fuel NDA Technique

Many different nondestructive analysis techniques are currently being investigated as a part of the United States Department of Energy's Next Generation Safeguards Initiative (NGSI) seeking methods to quantify plutonium in spent fuel. Neutron Resonance Transmission Analysis (NRTA) is one of these techniques. Having first been explored in the mid-1970s for the analysis of individual spent-fuel pins a second look, using advanced simulation and modeling methods, is now underway to investigate the suitability of the NRTA technique for assaying complete spent nuclear fuel assemblies. The technique is similar to neutron time-of-flight methods used for cross-section determinations but operates over only the narrow 0.1-20 eV range where strong, distinguishable resonances exist for both the plutonium (239, 240, 241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Initial modeling shows excellent agreement with previously published experimental data for measurements of individual spent-fuel pins where plutonium assays were demonstrated to have a precision of 2-4%. Within the simulation and modeling analyses of this project scoping studies have explored fourteen different aspects of the technique including the neutron source, drift tube configurations, and gross neutron transmission as well as the impacts of fuel burn up, cooling time, and fission-product interferences. These results show that NRTA may be a very capable experimental technique for spent-fuel assay measurements. The results suggest sufficient transmission strength and signal differentiability is possible for assays through up to 8 pins. For an 8-pin assay (looking at an assembly diagonally), 64% of the pins in a typical 17 ? 17 array of a pressurized water reactor fuel assembly can be part of a complete transmission assay measurement with high precision. Analysis of rows with up to 12 pins may also be feasible but with diminished precision. Preliminary data analysis of an NRTA simulation has demonstrated the simplicity of the technique.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗