William B. Wilson

Aggregate delayed neutron intensities and spectra using augmented ENDF/B-V precursor data

The evaluated nuclear data files (ENDF/B-V) have been used for 105 precursors, augmented with experimental spectral data for 29 emitters and recent model calculations for the remaining 76 emitters, in aggregate summation calculations. The equilibrium-delayed neutron intensities and spectra in the conventional six time groups for 11 fissionable nuclides were calculated. Normalized experimental neutron spectra for 29 emitters were supplied by Rudstam. Augmentation of the spectra uses data calculated with a recently developed code, BETA. Fission yields of the precursors and their delayed neutron branching fractions, P /SUB n/ , are from ENDF/B-V. Results are compared where possible with corresponding group and total evaluations of aggregate measurements. The intent of these calculations is to examine the adequacy of precursor data for inclusion in ENDF/B-VI. Most of the results, particularly with the added model parameters, are sufficiently accurate to be of direct interest to the reactor community. The total delayed neutron yield per neutron absorption, v /SUB d/ , for /sup 238/U and /sup 232/Th suggests that either an improvement in ENDF/B fission yields is needed or the evaluated experimental v /SUB d/ values are in significant error.

Operator declaration verification technique for spent fuel at reprocessing facilities

Abstract A verification technique for use at reprocessing facilities, which integrates existing technologies to strengthen safeguards through the use of environmental monitoring, has been developed at Los Alamos National Laboratory. This technique involves the measurement of isotopic ratios of stable noble fission gases from on-stack emissions during reprocessing of spent fuel using high-precision mass spectrometry. These results are then compared to a database of calculated isotopic ratios using a data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, reactor type, etc.). These inferred parameters can be used to verify operator declarations. The integrated system (mass spectrometry, reactor modeling, and data analysis) has been validated using on-stack measurements during reprocessing of fuel from a US production reactor. These measurements led to an inferred burnup that matched the declared burnup to within 3.9%, suggesting that the current system is sufficient for most safeguards applications. Partial system validation using gas samples from literature measurements of power reactor fuel has been reported elsewhere. This has shown that the technique developed here may have some difficulty distinguishing pressurized water reactor (PWR) from boiling water reactor (BWR) fuel; however, it consistently can distinguish light water reactor (either PWR or BWR) fuels from other reactor fuel types. Future validations will include advanced power reactor fuels (such as breeder reactor fuels) and research reactor fuels as samples become available.

Photonuclear Physics in Radiation Transport - III: Actinide Cross Sections and Spectra

Abstract This paper describes model calculations and nuclear data evaluations of photonuclear reactions on actinides such as 235U, 238U, 237Np, and 239Pu for incident photon energies from the reaction threshold up to 20 MeV. The calculations are done using the GNASH code, including the giant-dipole resonance for photoabsorption. The emission of secondary particles is computed using a preequilibrium theory, together with an open-ended sequence of the compound nucleus decay using the Hauser-Feschbach theory. The accuracy of the calculated and evaluated cross sections is assessed through extensive comparison with measured cross sections. This work also summarizes evaluation methods used to create actinide photonuclear files for the forthcoming ENDF/B-VII database, which will facilitate radiation transport studies related to photonuclear reactions in a number of technologies including production of photoneutrons and photofission fragments in electron accelerators, shielding studies, and nondestructive detection of nuclear material in particular.

Detection of antineutrinos for nonproliferation

Abstract The feasibility of using the detection of electron antineutrinos produced in fission to monitor the time dependence of the plutonium content of nuclear power reactors is discussed. If practical, such a scheme would allow worldwide, automated monitoring of reactors and, thereby, the detection of certain proliferation scenarios. For GW(electric) power reactors, the count rates and the sensitivity of the antineutrino spectrum (to the core burnup) suggest that monitoring of the gross operational status of the reactor from outside the containment vessel is feasible. As the plutonium content builds up in a given burn cycle, the total number of antineutrinos steadily drops; and this variation is quite detectable, assuming fixed reactor power. The average antineutrino energy also steadily drops, and a measurement of this variation would be very useful to help offset uncertainties in the total reactor power. However, the expected change in the antineutrino signal from the diversion of a significant quantity of plutonium, which would typically require the diversion of as little as a single fuel assembly in a GW(electric) reactor, would be very difficult to detect.

Noble‐gas atmospheric monitoring for international safeguards at reprocessing facilities

Environmental monitoring of nuclear activities promises to play a large role in the improvements in international safeguards under the International Atomic Energy Agencys Programme 93+2. Monitoring of stable noble‐gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr and Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CANDU fuel with a burnup of 1 GWd/tU. For each case, a maximum‐likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

Benchmarking of the Los Alamos neutron production rate code SOURCES-3A

Abstract The neutron production rate code SOURCES-3A was benchmarked using various experimental measurements from the literature. These experiments included thick-target yield measurements from Li, Be, B, C, O, F, Mg, Al, Si, UO 2 , and UC. Several integrated experiments (PuBe 13 and UO 2 F 2 homogeneous problems, Po–B 4 C and Po–Be interface problems, and Al 2 O 3 and SiO 2 beam problems) were also modeled, testing all the geometry characteristics of the SOURCES-3A code. The SOURCES-3A calculations were compared with the experimental results and good agreement was found in all cases. These benchmarks have shown that SOURCES-3A spectra and magnitude calculations are accurate to within ±18% for even the most complex problems.

Delayed-Gamma Simulation Using MCNPX

Abstract Monitoring issues related to activation and fission processes occur in many health physics, instrumentation and equipment design, nuclear forensics, and homeland security applications. Gamma radiation that is emitted during these processes as a result of the radioactive decay of reaction by-products [delayed gammas (DGs)] provides unique signatures that are useful for interrogation and information acquisition. Thus, it is of compelling interest to have a simulation tool that can be used to conduct studies to provide insights into the activation and fission processes. Beginning with version 2.5.0, MCNPX has been undergoing major upgrades to facilitate DG simulations. We illustrate the upgrades for a simple multiparticle reaction model involving 60Ni and for 235U photofission caused by 12-MeV photons.

Antineutrino monitoring of burning mixed oxide plutonium fuels

This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

DANDE-A Linked General Code System for Core Neutronics/Depletion Analysis

DANDE—a modular neutronics, depletion code system for reactor analysis—is described. It consists of nuclear data processing, core physics, and fuel depletion modules, and it allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of nuclide inventories and aggregate properties of burnable poisons, fission products, and actinides in the fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission product creation and decay, as well as inventories of higher order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated by two actual problems.

Compact Free-Electron Laser at the Los Alamos National Laboratory*

The design and construction of a second-generation free-electron laser (FEL) system at Los Alamos is described. Comprising state-of-the-art components, this FEL system will be sufficiently compact, robust, and user-friendly for application in industry, medicine, and research.