Simulations and experiments have been carried out to investigate the neutron emission characteristics of two mixed-oxide (MOX) fuels at Idaho National Laboratory (INL). These activities are part of a project studying advanced instrumentation techniques in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and its Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. This analysis used the MCNP-PoliMi Monte Carlo simulation tool to determine the relative strength and energy spectra of the different neutron source terms within these fuels, and then used this data to simulate the detection and measurement of these emissions using an array of liquid scintillator neutron spectrometers. These calculations accounted for neutrons generated from the spontaneous fission of the actinides in the MOX fuel as well as neutrons created via (D,n) reactions with oxygen in the MOX fuel. The analysis was carried out to allow for characterization of both neutron energy as well as neutron coincidences between multiple detectors. Coincidences between prompt gamma rays and neutrons were also analyzed. Experiments were performed at INL with the same materials used in the simulations to benchmark and begin validation tests of the simulations. Data was collected in these experiments using an array of four liquid scintillators and a high-speed waveform digitizer. Advanced digital pulse-shape discrimination algorithms were developed and used to collect this data. Results of the simulation and modeling studies are presented together with preliminary results from the experimental campaign. IntroductionAdvanced nuclear fuels are currently under development within the Department of Energy's Fuel Cycle Research and Development program as part of a long-term research effort focused at understanding the behavior of mixed-oxide (MOX) fuels containing minor actinides and long-lived fission products. The aim of this work is to understand how these materials impact the long-term performance of nuclear fuel in order to be able to design and manufacture advanced fuels for use in next-generation reactors. Reusing, or recycling, the higher actinides and long-lived fission products in advanced nuclear fuels ultimately leads to the transmutation of these materials into shorter-lived waste products which may be more easily and more safely disposed of. There are several potential benefits of reusing nuclear fuel including the reclamation of additional energy content from once-through used fuels, the reduction or removal of longer-lived waste products from spent fuel, and the lessening of the storage demands eventually placed on facilities for the long-term storage or disposal of spent fuels. In parallel with the fuel 5 (18) development projects research and development is also underway to develop advanced fuel reprocessing approaches to produce these fuels and to develop advanced reactors to use these fuels. However, in addition to these core engineering research and development projects ...

Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure ''spike'' solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137 Cs/ 137 Ba ratio measurements for ''age'' determination of radioactive 137 Cs sources, Idaho National Laboratory (INL) is producing enriched 134 Ba isotopes that are tobe used for IDMS spikes to accurately determine 137 Ba accumulation from the decay of 137 Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.