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The quantification of plutonium (Pu) in spent nuclear fuel is an increasingly important safeguards issue. There exists an estimated worldwide 980 metric tons of Pu in the nuclear fuel cycle and the majority is in spent nuclear fuel waiting for long term storage or fuel reprocessing. This study investigates utilizing the measurement of x-ray fluorescence (XRF) from the spent fuel for the quantification of its uranium (U) to Pu ratio. Pu quantification measurements at the front end of the reprocessing plant, the fuel cycle area of interest, would improve input accountability and shipper/receiver differences. XRF measurements were made on individual PWR fuel rods with varying fuel ages and final burn-ups at Oak Ridge National Laboratory (ORNL) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only been measured for fast breeder reactor fuel (~40 percent Pu). To understand the physics of the measurements, several modern physics simulations were conducted to determine the fuel isotopics, the sources of XRF in the spent fuel, and the sources of Compton continuum. Fuel transformation and decay simulations demonstrated the Pu/U measured peak ratio is directly proportional to the Pu/U content and increases linearly as burn-up increases. Spent fuel source simulations showed for 4 to 13 year old PWR fuel with burn-up ranges from 50 to 67 GWd/MTU, initial photon sources and resulting Compton and XRF interactions adequately model the spent fuel measured spectrum and background. The detector simulations also showed the contributions to the Compton continuum from strongest to weakest are as follows: the fuel, the shipping tube, the cladding, the detector can, the detector crystal and the collimator end. The detector simulations showed the relationship between the Pu/U peak ratio and fuel burn-up over predict the measured Pu/U peak but the trend is the same. In conclusion, the spent fuel simulations using modern radiation transport physics codes can model the actual spent fuel measurements but need to be benchmarked.

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (α,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

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This project aims to investigate and analyze crucial aspects related to the deployment of Small Modular Reactors (SMRs) in Sweden. SMRs have risen as a competitor to large-scale nuclear power plants with cost-competitiveness and cogeneration purposes in focus. Alongside the advantages of SMRs, it is important to address the non-proliferation and safeguards aspects of the deployment of SMRs in Sweden. The first stage of this project aimed to predict the electricity consumption in Sweden in 2045. Based on scenarios made by researchers and authorities, a prediction was made that the electricity consumption in Sweden will be approximately 240 TWh in 2045. Subsequently, a study was conducted to assess the suitability of different SMR concepts for deployment in Sweden. The results were that the GE Hitachi BWRX-300 and the Rolls Royce SMR are the concepts most suitable for electricity generation and the VTT LDR 50-SMR is suitable for district heating in Sweden. Based on the predicted electricity consumption and the predicted need for district heating, the SMRs were deployed in three different scenarios which yield the following results: In scenario i), a total of 15-23 SMRs were deployed to the existing nuclear sites in Sweden. (15 SMRs if BWRX-300 is chosen and 23 SMRs if the Rolls Royce SMR is chosen). In scenario ii), 16-24 SMRs in total were deployed to bidding area 1 (SE1), SE2, and SE4. In scenario iii), a total of 25 new reactors were deployed to eight different sites, close to where the electricity consumption was predicted to be large. The implications on material accounting and safeguards from the deployment in the scenarios were that the transportation and storage of spent nuclear fuel will rise by roughly a factor of two, depending on which scenario is chosen and the choice of SMR. There will be more sites for inspectors to visit. New harbors need to be built and licensed. New transportation infrastructure may need to be developed if scenario three is implemented. Furthermore, a new intermediate storage of spent nuclear fuel may be needed.

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The outdated and oversized nuclear manufacturing complex within the United States requires its transformation into a smaller, safe, and secure enterprise. Health and safety risks, environmental concerns, and the end of the Cold War have all contributed to this necessity. The events of September 11, 2001, emphasized the protection requirements for nuclear materials within the U.S. as well as abroad. Current Nuclear Safeguards regulations contain minimal prescriptive requirements relating to the design of new production facilities. Project management and engineering design guides require that design documents contain specific and measureable statements relating to systems requirements. The systems engineering process evaluates alternatives for an effective and integrated solution during project design. A Safeguards Design Strategy for domestic nuclear materials processing facilities based upon a core "framework" of safeguards regulatory programmatic elements that also use the prescriptive requirements and similar goals of safety, health, and physical security regulations is proposed and justifiable.

<p>Nuclear energy is currently one of the world’s main sources of electricity. Closely connected to the use of nuclear energy are important issues such as the nonproliferation of fissile material that may potentially used in nuclear weapons (safeguards), and the management of the highly radioactive nuclear waste. This thesis addresses both these issues by contributing to the development of new experimental methods for ensuring safe and secure handling of the waste, with focus on methods to be used prior to encapsulation and final storage.</p><p>The methods rely on high resolution gamma ray spectroscopy (HRGS), involving the measurement and analysis of emitted gamma radiation from the fission products ¹³⁷Cs, ¹³⁴Cs and ¹⁵⁴Eu. This technique is nondestructive, making it relatively nonintrusive with respect to the normal operation of the nuclear facilities.</p><p>For the safeguards issue, it is important to experimentally verify the presence and identity of nuclear fuel assemblies and also that the fuel has experienced normal, civilian reactor operation. It has been shown in this thesis that the HRGS method may be used for verifying operator declared fuel parameters such as burnup, cooling time and irradiation history. In the experimental part of the work, the burnup and the cooling time has been determined with an accuracy of 1.6% and 1.5%, respectively (1 σ).</p><p>A technique has also been demonstrated, utilizing the ratio ¹³⁴Cs/¹⁵⁴Eu, with which it is possible to determine whether a fuel assembly is of MOX or LEU type. This is of interest for safeguards as well as for the safe operation of a final storage facility.</p><p>As an improvement to the HRGS technique, measuring a part of the fuel assembly length in order to reduce measurement time has been suggested and investigated. A theoretical case for partial defect verification has also been studied as an extension of the HRGS technique. </p><p>Finally, HRGS has been used for determining the decay heat in spent nuclear fuel assemblies, which is of importance for the safe operation of a final storage facility. This application is based on the radiation from ¹³⁷Cs, and the accuracy demonstrated was within 3% (1 σ).</p>