Neutron coincidence counting is an important passive Nondestructive Assay (NDA) technique widely used for qualitative and quantitative analysis of nuclear material in bulk samples. During the fission process, multiple neutrons are simultaneously emitted from the splitting nucleus. These neutron groups are often referred to as coincident neutrons. Because different isotopes possess different coincident neutron characteristics, the coincident neutron signature can be used to identify and quantify a given material. In an effort to identify unknown nuclear samples in field inspections, the Portable Neutron Coincidence Counter (PNCC) has been developed. This detector makes use of the coincident neutrons being emitted from a bulk sample. An in-depth analysis has been performed to establish whether the nuclear material in an unknown sample could be quantified with the accuracy and precision needed for safeguards measurements. The analysis was performed by comparing experimental measurements of PuO2 samples to the calculated output produced using MCNPX and the Neutron Coincidence Point Model. Based on the analysis, it is evident that this new portable system can play a useful role in identifying nuclear material for verification purposes.

Thesis (Ph.D.)--University of California, Berkeley, 1961. / "Chemistry, UC-4" -t.p. "TID-4500 (16th Ed.)" -t.p.

This work considers the impact of fast reactor fuel cycles on the UK's nuclear waste inventory, focusing on the disposition of the UK's plutonium stockpile and spent fuel from new build nuclear reactors. Reprocessing spent fuel from nuclear reactors has led to a large stockpile of civil plutonium in the UK. At the end of reprocessing the stockpile was estimated to be 112 tonnes. This large stockpile of separated plutonium poses a proliferation concern and there is no strategy at present for UK plutonium disposition. The NDA's position paper in 2014 stated the re-use of plutonium in a reactor as a preferred option. These options included Mixed OXide (MOX) fuelled Pressurised Water Reactors (PWR) and the use of plutonium in a Sodium-cooled Fast Reactor (SFR), PRISM, operated as a once-through plutonium burning fast reactor. As yet a preferred option has not been selected by the government. Nuclear power is the UK's largest source of low-carbon electricity. Current plans aim to build 16 GWe of new reactors by 2050 to replace the UK's current fleet. This work considered PWR MOX and once-through SFRs for UK plutonium disposition, comparing their relative merits to the direct disposal of the plutonium stockpile in a geological repository. The waste performance of disposition options were compared using assessment criteria based on: Technology Readiness Level (TRL), final stockpile mass, repository size and radiotoxicity. To maximise the reduction of the UK's plutonium stockpile, closed SFR fuel cycles were also considered with scenarios aimed at improving waste performance. Once-through and closed SFR fuel cycles were also considered for the disposition of spent fuel from new build reactors. Research presented in this thesis shows that UK waste disposition options are highly dependent on fuel cycle operating parameters. In once-through plutonium disposition options all scenarios increased repository size compared to direct disposal. Once-though SFRs increased repository size the least, where as PWR MOX reduced the stockpile mass most significantly. The most significant improvement in waste performance, using a closed fuel cycle up to 2150, required short reprocessing times and americium reprocessing. There were no additional improvements of significance with curium reprocessing and the choice of metallic or MOX fuelled SFRs had little impact on waste performance. Preferred fuel cycle scenarios are dependent on the priority given to different assessment criteria. To compare fuel cycle scenarios on an even basis, decision analysis methods were presented using assessment criteria results from the fuel cycles modelled in this work. Decision analysis methods were designed so that the reader can apply their own priorities, through the use of weightings, to the assessment criteria to determine preferable fuel cycle scenarios.

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This research uses an existing innovative fuel design (IFD) that has intrinsic safety features and enhanced economics over the current uranium dioxide (UO2) light water fuel design and evaluates promising methods to improve the waste management and proliferation resistance of the IFD by doping the fresh fuel with select actinides.The most robust approach for proliferation resistance is to denature these materials by adding a uranium or plutonium isotope that hampers the usability of the materials in weapons. The proposed modifications to the IFD use this approach through elevated fractions of 238Pu. 238Pu generates large quantities of heat and neutrons through its radioactive decay and is estimated to make plutonium potentially “proliferation-proof." The IFD this work uses as a foundation is an advanced metallic fuel designed for use in current light water reactors. Due to the high fission density of metallic fuel and the proposed uranium enrichments, the plutonium produced by irradiating this fuel has promising isotopic content for proliferation resistance. This proliferation resistance will be further increased by adding 237Np and/or 241Am to the initial fresh fuel composition that will result in increased 238Pu content. Adding these actinides into the fresh fuel at 0.2 wt.%, the amount of 238Pu produced in the used fuel can be used for proliferation resistance. Increasing the actinide wt.% can potentially produce "proliferation-proof" used fuel. Also, by utilizing neptunium and americium in fresh fuel, many of the challenges with permanent geological disposal of used fuel can be mitigated.

When an accident involving the possibility of a plutonium contaminated wound occurs, the contamination is often quantified using sodium iodide (NaI(Tl)) and high purity germanium (HPGe) detection systems. The NaI(Tl) system is used to quantify the amount of contamination, while HPGe is used to gauge the depth of contamination in the wound. Assessment of plutonium contaminated wounds is difficult due to the lowenergy and yield of the uranium L-shell x rays used for the measurement, which can be effected by source distance, shape, and tissue attenuation. These effects on wound counting systems used at Los Alamos National Laboratory (LANL) were characterized experimentally using common source shapes (disk, point, and line) and acrylic plastic as a tissue substitute. Experiments were conducted to characterize detector responses as a function of tissue attenuation, source distance, and source depth in tissue. The computer code MCNP5 was used to model both systems for wound counting and better examine angular displacement of a line source in tissue. The NaI(Tl) detector response was characterized using absolute detector efficiency for all experimental measurements. Measurements showed that the NaI(Tl) system is significantly effected by the source to detector position and depth in tissue. Characterization of the HPGe detection system was done utilizing the peak-to-peak ratio from the two low-energy x rays. HPGe peak-to-peak ratios were not affected by source to detector distance, but showed an increased response to source depth in tissue. MCNP results suggested that small incident angles from the plane of the detector face can cause significant effects on the response of both detectors. In summary, the response of both systems showed dependence on source geometry and depth of contamination in tissue. Correction values and uncertainties were determined based on these dependencies.

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: 1) SINRD provides absolute measurements of burnup independent of the operator’s declaration. 2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3σ from LWR spent LEU and MOX fuel. 3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. 4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. 5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

Gloveboxes, where special nuclear material is handled and such as those present at Los Alamos National Labs, LANL, provide an experimental area confined within a protective shell and with strict environmental controls. These gloveboxes allow workers to indirectly interact with hazardous material. Unfortunately, these gloveboxes are not fail proof and are subject to occasional accidental failures resulting in possible breaches of containment and release of nuclear material. In particular, fires within the gloveboxes are of major concern with regard to the potential for breaches and damage to not only the glovebox but also to surrounding areas as well. Another, potentially even catastrophic, result of glovebox fires is the potential for the spread of radioactive contamination. There is some historical precedent of contaminant release resulting from glovebox fires, such as those at the Rocky Flats Plant (Buffer, 2012). Gloveboxes at LANL are currently equipped with manually activated fire suppression systems. In the event of an incident, a worker would hit a nearby emergency button and the system would be activated. However, this method relies on the worker to have the presence of mind in the face of danger to activate the system, and as such there is no true guarantee that the systems will be triggered. Since the level of consequence is dire, then the ideal situation requires that other fire suppression systems be present which do not rely on human interaction to function. The Fire Foe™ system has been chosen as a secondary failsafe measure in order to meet this need. Analysis of how the casing of the Fire Foe™ system, composed of nylon 6,6 polymer, weathers under irradiation in gloveboxes is paramount in determining the effectiveness and potential lifetimes of the systems within the gloveboxes. Samples of nylon 6,6 were exposed to a 5 Ci PuBe neutron source located at the University of Texas as well as a high dose rate beam of 4.5 MeV alpha particles located at Los Alamos to determine the effect of neutron and alpha particle damage on the polymer material. Subsequent mechanical testing was conducted to determine alteration to the tensile properties of the nylon 6,6 material for both irradiated and non-irradiated samples. / text

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