As the nuclear industry grows around the globe, it brings with it a need for more safeguards and proliferation resistant technologies. The International Atomic Energy Agency (IAEA) depends on effective containment and surveillance (C/S) technologies and methods for maintaining continuity of knowledge over nuclear assets. Tags and seals, a subset of C/S technologies, are an area where innovation has been relatively stagnant for the past fifteen years. It is necessary to investigate technologies not previously used in this field in order to defend against emerging threats and methods of defeat. Based on a gap analysis of tags and seals currently being used by the IAEA, completed with the input of several subject matter experts, the technology selected for investigation was ion-exchanged glass. Ion-exchanged glass is relatively inexpensive, has high strength, and can be used in a variety of applications. If identical pieces of glass are exchanged under the same conditions and subjected to the same point load, the fracture patterns produced can be compared and used as a verification measure. This technology has the potential to be used in passive seal applications. Each image was categorized depending on its fracture as a "3 leaf" or "4 leaf" pattern. These two populations were separately analyzed and evaluated. Several methods used to analyze the fracture patterns involve the use of image analysis software such as ImageJ and the MATLAB Control Point Selection Tool. The statistical analysis software Minitab was used to validate the use of facture pattern analysis as verification tool. The analysis yielded a 60% verified comparison for samples demonstrating a "3 leaf" fracture pattern and a 78% verified comparison for samples with a "4 leaf" fracture pattern. This preliminary analysis provides a strong indication of the plausibility for the use of ion-exchanged glass as a verification measure for C/S measures and specifically tags and seals.

Nuclear reactor safeguards are how the peaceful use of nuclear material is ensured. Safeguards consist of a broad array of techniques, such as video surveillance and tamperproof seals, to ensure that nuclear material is not diverted from declared activities. Safeguards research is conducted to ensure that safeguards techniques are capable of meeting the challenges posed by future reactor designs and operating conditions. One such technique that has broad applicability to novel reactor designs, including small modular reactors, is the method of standoff neutron detection using large area neutron detectors. In this method, neutron detectors are employed to detect neutrons which have escaped from the core, which are representative of the flux inside the core. Because the flux required to achieve a given power is dependent on the isotopes being fissioned, due to their different cross sections and fission energies, the state of the core can be assessed using the neutron detectors. Prior research has demonstrated that it is possible to correlate kilogram changes in fissile inventory using neutron detectors by employing the standoff neutron detector method. This work at the McMaster nuclear reactor details additional experiments to support prior work. First, the apparatus and procedure to collect neutron detector data are detailed, along with persistent challenges to the collection. Next, simulations using the OSCAR-5 code to determine the fissile inventory are described. These two sections are subsequently combined, to compare changes in detector signal to the simulated core inventory. It was found that the uncertainty was too large to correlate changes in detector signal with changes in core inventory. Lastly, a method of detecting malicious interference is derived and tested. / Thesis / Master of Applied Science (MASc)

Since the first observation of antineutrinos from beta decay of the fission products inside a nuclear reactor in 1956, the design and operating experience of antineutrino detectors near reactors has increased to the point where monitoring the reactor's power level and progression through its burnup cycle has become possible. With the expected increase in world nuclear energy capacity, including the dissemination of reactor technologies to non-nuclear states, the need for safeguards measures which are able to provide continuous, near-real-time information about the state of the core, including its isotopic composition, in a tamper- and spoof-resistant manner is evident. Near-field (~20 m from the core) antineutrino detectors are able to fulfill this demand without perturbing normal reactor operation, without requiring instrumentation which penetrates the reactor vessel, and without displacing other plant structures. Two sodium-cooled long-life fast reactors that are characteristic of next-generation reactors which are attractive for installation in non-nuclear states, one large and one small power rating, have been modeled throughout their reference burnup cycles using MCC-3 and DIF3D/REBUS. Various diversions of fissile material from the core designed to obtain weapons-usable material for the purpose of nuclear proliferation were studied as perturbed core states. The difference in detector event rates between the reference and perturbed states was used to determine the probability that a particular diversionary activity would be apparent before the material could be converted into a weapon. These data indicate which types of diversion antineutrino safeguards are particularly strong against and how the technology might be implemented in current and future international policies concerning nuclear proliferation.

There has been a strong international interest in using pyroprocessing to close the fast nuclear reactor fuel cycle and reprocess spent fuel efficiently. To commercialize pyroprocessing, safeguards technologies are required to be developed. In this research, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been investigated as a method to safeguard the process and more precisely quantify the 239Pu content of pyroprocessing materials. This method uses a detector array with different filters to isolate the low-energy resonance in 239Pu neutron fission cross section. The relative response of the different detectors allows for the quantification of the amount of 239Pu in the pyroprocessing materials. The Monte-Carlo N-Particle (MCNP) code was used to design a prototype SINRD instrument. This instrument is composed of a neutron source pod and a SINRD detector pod. Experimental measurements were also performed to validate the MCNP model of the instrument. Based on the results from simulations and experiments, it has been concluded that the MCNP model accurately represents the physics of the experiment. In addition, different SINRD signatures were compared to identify which of them are usable to determine the fissile isotope content. Comparison of different signatures allowed for reduction in the uncertainty of the 239Pu mass estimate. Using these signatures, the SINRD instrument was shown to be able to quantify the 239Pu content of unknown pyroprocessing materials suitable for safeguards usage.

<p>Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented.</p><p>As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level.</p><p>The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities.</p><p>For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.</p>

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented. As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level. The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities. For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.

Pebble-fueled high temperature gas-cooled reactor (HTGR) technology was first developed by the Federal Republic of Germany in the 1950s. More recently, the design has been embraced by the People's Republic of China and the Republic of South Africa. Unlike light water reactors that generate heat from fuel assemblies comprised of fuel rods, pebble-fueled HTGRs utilize thousands of 60-mm diameter fuel spheres (pebbles) comprised of thousands of TRISO particles. As this reactor type is deployed across the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material safeguards measures to deter and detect the diversion of fuel pebbles. The disadvantages to these approaches are the loss of continuity of knowledge (CoK) when C/S systems fail, or are compromised, and the introduction of material unaccounted for (MUF). Either vulnerability can be exploited by an adversary to divert fuel pebbles from the reactor system. It was determined that a solution to maintaining CoK is to develop a system to identify each fuel pebble that is inserted and removed from the reactor. Work was performed to develop and evaluate the use of inert microspheres placed in each fuel pebble, whose random placement could be used as a fingerprint to identify the fuel pebble. Ultrasound imaging of 1 mm zirconium oxide microspheres was identified as a possible imaging system and microsphere material for the new safeguards system concept. The system concept was evaluated, and it was found that a minimum of three microspheres are necessary to create enough random fingerprints for 10,000,000 pebbles. It was also found that, over the lifetime of the reactor, less than 0.01% of fuel pebbles can be expected to have randomly the same microsphere fingerprint. From an MCNP 5.1 model, it was determined that less than fifty microspheres in each pebble will have no impact on the reactivity or temperature coefficient of reactivity of the reactor system. Finally, using an ultrasound system it was found that ultrasound waves can penetrate thin layers of graphite to image the microsphere fingerprint.

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.