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)

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.

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 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.

Nuclear safeguards are intrinsic and extrinsic features of a facility which reduce probability of the successful acquisition of special nuclear material (SNM) by hostile actors. Future bulk handling facilities in the United States will include both domestic and international safeguards as part of a voluntary agreement with the International Atomic Energy Agency. A new framework for safeguards, the Safeguards Envelope Methodology, is presented. A safeguards envelope is a set of operational and safeguards parameters that define a range, or “envelope,” of operating conditions that increases confidence as to the location and assay of nuclear material without increasing costs from security or safety. Facilities operating within safeguards envelopes developed by this methodology will operate with a higher confidence, a lower false alarm rate, and reduced safeguards impact on the operator. Creating a safeguards envelope requires bringing together security, safety, and safeguards best practices. This methodology is applied to an example facility, the Idaho Chemical Processing Plant. An example diversion scenario in the front-end of this nuclear reprocessing facility, using actual operating data, shows that the diversion could have been detected more easily by changing operational parameters, and these changed operational parameters would not sacrifice the operational efficiency of the facility, introduce security vulnerabilities, or create a safety hazard.

The Digital Cherenkov Viewing Device (DCVD) is an instrument used by authority inspectors to assess irradiated nuclear fuel assemblies in wet storage for the purpose of nuclear safeguards. Originally developed to verify the presence of fuel assemblies with long cooling times and low burnup, the DCVD accuracy is sufficient for partial defect verification, where one verifies that part of an assembly has not been diverted. Much of the recent research regarding the DCVD has been focused on improving its partial defect detection capabilities. The partial-defect analysis procedure currently used relies on comparisons between a predicted Cherenkov light intensity and the intensity measured with the DCVD. Enhanced prediction capabilities may thus lead to enhanced verification capabilities. Since the currently used prediction model is based on rudimentary correlations between the Cherenkov light intensity and the burnup and cooling time of the fuel assembly, there are reasons to develop alternative models taking more details into account to more accurately predict the Cherenkov light intensity. This work aims at increasing our understanding of the physical processes leading to the Cherenkov light production in irradiated nuclear fuel assemblies in water. This has been investigated through simulations, which in the future are planned to be complemented with measurements. The simulations performed reveal that the Cherenkov light production depends on fuel rod dimensions, source distribution in the rod and initial decay energy in a complex way, and that all these factors should be modelled to accurately predict the light intensity. The simulations also reveal that for long-cooled fuel, Y-90 beta-decays may contribute noticeably to the Cherenkov light intensity, a contribution which has not been considered before. A prediction model has been developed in this work taking fuel irradiation history, fuel geometry and Y-90 beta-decay into account. These predictions are more detailed than the predictions based on the currently used prediction model. The predictions with the new model can be done quickly enough that the method can be used in the field. The new model has been used during one verification campaign, and showed superior performance to the currently used prediction model. Using the currently used model for this verification, the difference between measured and predicted intensity had a standard deviation of 15.4% of the measured value, and using the new model this was reduced to 8.4%.

Two important issues related to nuclear materials safeguards are the continuous monitoring of nuclear processing facilities to verify that undeclared uranium is not processed or enriched and to verify that declared uranium is accounted for. The International Atomic Energy Agency (IAEA) is tasked with ensuring special nuclear facilities are operating as declared and that proper material safeguards have been followed. Traditional safeguards measures have relied on IAEA personnel inspecting each facility and verifying material with authenticated instrumentation. In newer facilities most plant instrumentation data are collected electronically and stored in a central computer. Facilities collect this information for a variety of reasons, most notably for process optimization and monitoring. The field of process monitoring has grown significantly over the past decades, and techniques have been developed to detect and identify changes and to improve reliability and safety. Several of these techniques can also be applied to international and domestic safeguards. This dissertation introduces a safeguards monitoring system developed for both a simulated Uranium blend down facility, and a water-processing facility at the Oak Ridge National Laboratory. For the simulated facility, a safeguards monitoring system is developed using an Auto-Associative Kernel Regression model, and the effects of incorporating facility specific radiation sensors and preprocessing the data are examined. The best safeguards model was able to detect diversions as small as 1.1%. For the ORNL facility, a load cell monitoring system was developed. This monitoring system provides an inspector with an efficient way to identify undeclared activity and to identify atypical facility operation, included diversions as small as 0.1 kg. The system also provides a foundation for an on-line safeguards monitoring approach where inspectors remotely facility data to draw safeguards conclusion, possibly reducing the needed frequency and duration of a traditional inspection.

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The safeguards verification of spent nuclear fuel is one of the major concern for the safeguards community, as this material represents about 80% of all material placed under safeguards.This PhD thesis described the development of two passive non-destructive assay (NDA) techniques: the Self-Indication Neutron Resonance Densitometry (SINRD) and the Partial Defect Tester (PDET).The NDA methods were investigated with Monte Carlo simulations and the benchmark experiments for SINRD were performed at the GELINA facility of JRC-IRMM in Geel, Belgium.The results for the SINRD technique showed promising results for the direct quantification of 239Pu in spent fuel, and both techniques gave encouraging results for the detection of partial defects. / Doctorat en Sciences de l'ingénieur et technologie / info:eu-repo/semantics/nonPublished