Accurate nuclear data provide an essential foundation for advances in a wide range of fields, including nuclear energy, nuclear safety and security, safeguards, nuclear medicine, and planetary and space exploration. In these and other critical domains, outdated, imprecise, and incomplete nuclear data can hinder progress, limit precision, and compromise safety. Similar nuclear data needs are shared by many applications, thus prioritizing these needs is especially important and urgently needed. Many levels of analysis are required to prepare nuclear measurements for employment in end-user applications. Because research expertise is typically limited to one level, collaboration across organizations and international borders is essential. This perspective piece provides the latest advances in nuclear data for applications and describes an outlook for both near- and long-term progress in the field.

Interest in safeguards verification measurements using passive thermal neutron counting to assay U-235 content in large 30B UF6 canisters has grown in recent years. The prohibitively high cost and impracticality of using reference 30B calibration cylinders extensively will likely make accurate simulations of increasing interest. Accuracy of the simulated response will define the confidence in the predicted response and the extent to which simulations can reasonably be relied upon. With U-234 driven F-19(alpha, n) reactions being the main neutron source in low enriched UF6 the uncertainties of the F-19(alpha, n) energy spectrum and the thick target yield of U-234 in UF6 propagate into the uncertainty in the predicted response and represent a major influence of basic nuclear data. Here sensitivity of the simulated total (Singles) and coincidence (Doubles) count rates are assessed for the Passive Neutron Enrichment Meter using six potential F-19(alpha, n) neutron energy spectra over a range of enrichments and material distributions. The results indicate that variations in the Singles and Doubles due to simulated (alpha, n) neutron spectrum are less than 1.5% for this set of simulated neutron spectra, with dependence varying inversely with enrichment. Singles uncertainty is only slightly less than that of the thick target F-19(alpha, n) yield corresponding to the primary neutron source, whereas the F-19(alpha, n) yield dependence of the Doubles is reduced by the non-negligible U-238(SF) coincident neutron emissions. Based on available thick target F-19(alpha, n) yield estimates the uncertainty is on the order of 5%, establishing this as the main nuclear data limitation when simulating thermal neutron detectors response for 30B UF6 storage cylinders. Based on these findings, it appears that the measurement and evaluation of the thick target F-19(alpha, n) yield for uranium hexafluoride is due.

Mobile radiation detector systems are important tools for detecting radiological and nuclear sources outside of regulatory control, but due to their mobility, they are subject to complex and varying backgrounds in most realistic operational scenarios. Recent work has found correlations between non-radiological contextual information and gamma-ray spectral features that can be used to decrease false alarm rates, however a more complete understanding of background source terms has been elusive. A measurement campaign developed a full radiological characterization of a controlled facility that roughly corresponded to two city blocks. As part of the campaign, the Radiological Multi-sensor Analysis Platform (RadMAP) collected extensive multi-sensor data. In this work, RadMAP's panoramic video data were used to visually identify several different materials and quantify the detector response to each material. A linear model was fit between the material responses and radiological features, and the results demonstrate reasonable agreement with ground truth flux measurements for each material. This paper will describe the data collection, processing, and analysis of the gamma-ray and video data from RadMAP. The paper will conclude with perspectives on the applicability of such a method to less controlled environments both with respect to achieving better understanding the sources of variability of background radiation in urban environments and whether such methods could be leveraged in operational scenarios.

A well-established knowledge of nuclear phenomena including fission, reaction cross sections, and structure/decay properties is critical for applications ranging from the design of new reactors to nonproliferation to the production of radioisotopes for the diagnosis and treatment of illness. However, the lack of a well-quantified, predictive theoretical capability means that most nuclear observables must be measured directly and used to calibrate empirical models, which in turn provide the data needed for these applications. In many cases, either there is a lack of data needed to guide the models or the results of the different measurements are discrepant, leading to the development of evaluation methodologies to provide recommended values and uncertainties. In this review, we describe the nuclear data evaluation process and the international community that carries it out. We then discuss new measurements and improved theory and/or modeling needed to address future challenges in applied nuclear science.

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 x 15 design with an initial enrichment of 4.013 wt% U-235, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via high performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL); and these data have been widely used to support code and nuclear data validation. The recent measurements performed by ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. These measurements serve to improve confidence in the data, to verify reported uncertainties, and to investigate previous anomalies noted in the plutonium measurements. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation. (C) 2015 Elsevier Ltd. All rights reserved.

A Lead Slowing Down Spectrometer (LSDS) provides a high neutron flux environment that enables measurements of small samples (similar to mu g) or samples with small cross sections (tens of mu b). The LSDS at Rensselaer Polytechnic Institute (RPI) was previously used for fission cross section measurements and for studies of methods for assay of used nuclear fuel. The effective energy range for the LSDS is 0.1 eV to 10 keV with energy resolution of about 35%. Two new LSDS applications were recently developed at RPI; the first enables simultaneous measurements of the fission cross section and fission fragment mass and energy distributions as a function of the incident neutron energy. The second enables measurements of the (n,alpha) and (n, p) cross sections for materials with a positive Q value for these reactions. Fission measurements of (252)Cf, (235)U, and (239)Pu were completed and provide information on fission fragment and energy distributions in resonance clusters. Measurements of the (n,alpha) cross section for (147,149)Sm were completed and compared to previously measured data. The new data indicate that the existing evaluations need to be adjusted.

This work describes a nondestructive strategy and algorithm designed to evaluate the burnup and plutonium content of light-water-reactor spent fuel and thereby addresses discrepancies with declared values. In contrast with previous methods that focus on only a few photopeaks (e.g., 137 Cs at 662 keV), the present approach involves the entire gamma spectrum up to 2000 keV. Spectra are used as input for the inverse code INDEPTH, which is designed to predict reactor parameters (fuel enrichment, power level, irradiation time, and cooling time) when given either a set of nuclide inventories or the gamma spectrum that they produce. This approach has the advantage of often making possible the determination of parameters other than burnup when they are unknown or in doubt. In addition, error in one photopeak evaluation is mitigated by the inclusion of the entire spectrum. The solution procedure involves multiple runs of the forward code ORIGEN/ARP, each of which produces an extensive list of nuclides formed through depletion/decay processes. The gamma spectrum of these nuclides is compared with the gamma spectrum from a detector through a bin-by-bin sum-of-squared-error. New choices for the reactor parameters that are input to ORIGEN/ARP are determined using a gradient search technique, and the best parameter set is that which minimizes the squared error between calculated and measured gamma spectra. The method is applied to the analysis of gamma data taken from various sections of actual spent reactor fuel and is compared with declared values and other methods of evaluation. The sensitivity of the inverse solution with respect to various parameters is calculated and indicates that the algorithm is stable and robust. One example includes the presence of multiple solutions, each of which can be characterized using additional information.

A new method of measuring fission fragment mass and energy distributions as a function of incident neutron energy in the range from below 0.1 eV to 1 keV has been developed. The method involves placing a double-sided Frisch-gridded fission chamber in Rensselaer Polytechnic Institute's lead slowing-down spectrometer (LSDS). The high neutron flux of the LSDS allows for the measurement of the energy-dependent, neutron-induced fission cross sections simultaneously with the mass and kinetic energy of the fission fragments of various small samples. The samples may be isotopes that are not available in large quantities (submicrograms) or with small fission cross sections (microbarns). The fission chamber consists of two anodes shielded by Frisch grids on either side of a single cathode. The sample is located in the center of the cathode and is made by depositing small amounts of actinides on very thin films. The chamber was successfully tested and calibrated using 0.41 +/- 0.04 ng of Cf-252 and the resulting mass distributions were compared to those of previous work. As a proof of concept, the chamber was placed in the LSDS to measure the neutron-induced fission cross section and fragment mass and energy distributions of 25.3 +/- 0.5 mu g of U-235. Changes in the mass distributions as a function of incident neutron energy are evident and are examined using the multimodal fission mode model.

A double sided Frisch-gridded fission chamber for use in RPI's lead slowing-down neutron spectrometer (LSDS) is being developed at Rennselaer Polytechnic Institute. Placing this fission chamber in the high neutron flux of the LSDS allows the measurement of the energy dependent, neutron induced fission cross sections, as well as the mass and kinetic energy of the fission fragments of various small samples. The fission chamber consists of two anodes shielded by Frisch grids on either side of a single cathode. The sample is located in the center of the cathode and is made by depositing small amounts of actinides dissolved in solution on very thin films. The chamber was successfully tested and calibrated using 0.4 ng of Cf-252 and mass distributions were compared to previous work. As a proof of concept, the chamber was then placed in the LSDS to measure, simultaneously, the neutron induced fission cross section and fragment mass and energy distributions of 24.9 mu g of U-235. The mass distribution as a function of neutron energy was examined and it may be possible to see changes in the mass distribution as a function of neutron energy. This process will enable the measurement of isotopes that are not available in large enough quantities (sub-micrograms) or with small fission cross sections (microbarns).

A Lead Slowing-Down Spectrometer (LSDS) was recently installed at LANL [D. Rochman, R.C. Haight, J.M. O'Donnell, A. Michaudon, S.A. Wender, D.J. Vieira, E.M. Bond, T.A. Bredeweg, A. Kronenberg, J.B. Wilhelmy, T. Ethvignot, T. Granier, M. Petit, Y. Danon, Characteristics of a lead slowing-down spectrometer coupled to the LANSCE accelerator, Nucl. Instr. and Meth. A 550 (2005) 397]. The LSDS is comprised of a cube of pure lead 1.2 m on the side, with a spallation pulsed neutron source in its center. The LSDS is driven by 800 MeV protons with a time-averaged current of up to 1 mu A, pulse widths of 0.05-0.25 mu s and a repetition rate of 20-40 Hz. Spallation neutrons are created by directing the proton beam into an air-cooled tungsten target in the center of the lead cube. The neutrons slow down by scattering interactions with the lead and thus enable measurements of neutron-induced reaction rates as a function of the slowing-down time, which correlates to neutron energy. The advantage of an LSDS as a neutron spectrometer is that the neutron flux is 3-4 orders of magnitude higher than a standard time-of-flight experiment at the equivalent flight path, 5.6 m. The effective energy range is 0.1 eV to 100 keV with a typical energy resolution of 30% from 1 eV to 10 keV The average neutron flux between 1 and 10 keV is about 1.7 x 10(9) n/cm(2)/s/mu A. This high flux makes the LSDS an important tool for neutron-induced cross section measurements of ultra-small samples (nanograms) or of samples with very low cross sections. The LSDS at LANL was initially built in order to measure the fission cross section of the short-lived metastable isotope of U-235, however it can also be used to measure (n,alpha) and (n,p) reactions. Fission cross section measurements were made with samples of U-235, U-236, U-238 and Pu-239. The smallest sample measured was 10 ng of Pu-239. Measurement of (n,alpha) cross section with 760 ng of Li-6 was also demonstrated. Possible future cross section measurements include fission and (n,p) and (n,alpha) reaction in radioactive samples.(c) 2007 Elsevier B.V. All rights reserved.

We present fission cross-section measurements with ∼10ng of 239Pu performed using the LANSCE Lead Slowing-Down Spectrometer. Results of Li6(n,α) measurements with a sample size of 760ng of 6Li are also reported. This technical achievement demonstrates the feasibility of measuring neutron-induced fission cross-section on samples with 10ng of fissile actinides that are available on ultra-small quantities. Furthermore, results on neutron-induced alpha emission show that measurements for astrophysics purposes are feasible with the LSDS.

At the Los Alamos Neutron Science Center (LANSCE) a compensated ionization chamber (CIC) was placed in a lead slowing down spectrometer (LSDS) to measure the 6Li(n,α)3H cross-section as a feasibility test for further work. The LSDS consists of a 1.2m cube of lead with a tungsten target in the center where spallation neutrons are produced when bombarded with pulses of 800MeV protons. The resulting neutron flux is of the order of 1014n/cm2/s which allows the cross-section measurement of samples of the order of 10's of nanograms. The initial experiment measured a 91μg sample of natural lithium flouride. Cross-section measurements were obtained in the 0.1eV–2keV energy range. A 62μg sample was placed in the chamber with a higher neutron beam intensity, and data was obtained in the 0.1–300eV range. Adjustments in chamber dimensions and electronic configuration will improve gamma flash compensation at high beam intensity, decrease the dead time, and increase the energy range where data can be obtained. The intense neutron flux will allow the use of a smaller sample.