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Dry casks will be a prevalent spent nuclear fuel (SNF) storage option until solutions for long-term storage or disposal are deployed. A dry cask storing 32 pressurized water reactor fuel assemblies will likely contain about 20 significant quantities of plutonium, so these structures require effective safeguards monitoring. An external remote monitoring system (RMS) is proposed to advance the current dry cask safeguards regime which relies on containment and surveillance. The objectives of this study were to assess the performance of the external RMS as a detection system and to develop a simulation approach for estimating measurements. Small-scale experiments of generic neutron source diversions mimicking SNF diversion from a dry cask were conducted and the nondetection probability was calculated for a variety of measurement times. MCNP simulations were carried out to assess the degree to which the measurement results could be predicted. A previous simulation methodology was advanced to consider uncertainty in the activity of sources being measured. The study concluded that the external RMS performs well as a neutron detection system and that MCNP simulation is a viable tool both for predicting measurements made with the external RMS and for calculating nondetection probabilities of hypothetical, generic diversion scenarios. [ABSTRACT FROM AUTHOR]

It has become a common practice to store sufficiently cooled spent nuclear fuel (SNF) assemblies in interim storage dry casks with passive cooling. These dry casks require nuclear safeguards monitoring because they contain plutonium. Past studies on dry cask modeling and simulations have shown that a remote monitoring system (RMS) situated inside the dry cask could continually monitor and detect the removal of even a single SNF assembly from the cask. This conceptual RMS design was tested by conducting laboratory-scale experiments using small-size 252Cf neutron sources. These small-size sources were surrounded by neutron-reflecting materials in the experiments to mimic the SNF assemblies as a surface neutron source to the fission chamber detectors of the RMS. Experimental and simulation results showed that the removal or diversion of even a single neutron source is detectable within 4 min with a probability of detection greater than 80%. [ABSTRACT FROM AUTHOR]

A nuclear forensics technique, based on the maximum likelihood method, for the attribution of reactor type, fuel burnup, and time since irradiation (TSI) of separated pure plutonium (Pu) samples was previously developed at Texas A&M University. The method utilized measured values of ten intra-elemental isotope ratios in the Pu sample and a large database consisting of the values for these ratios as a function of the three attributes: reactor type, fuel burnup, and TSI. However, this method failed for Pu samples with mixed attributes. Hence, a new technique based on machine learning methods was developed that matched the capabilities of the previous maximum likelihood method for pure Pu samples. This new methodology used support vector machines for reactor-type discrimination and Gaussian process regression for fuel burnup quantification. The TSI was calculated analytically using the predicted reactor type and fuel burnup. This new method holds great potential for the attribution of mixed Pu samples. [ABSTRACT FROM AUTHOR]

Until a long-term solution is finalized, interim storage of sufficiently cooled spent nuclear fuel (SNF) assemblies in dry casks is the predominant practice. Since these dry casks can contain approximately 160 kg of reactor-grade plutonium, they require safeguards monitoring. Results of a simulation study conducted on the design development and analysis of a remote monitoring system (RMS) are presented. The goal of the study was to determine the suitability of this RMS to meet the SNF monitoring objectives. MCNP simulations of a dry cask with all its contents and a set of simulations with one or two removed SNF assemblies were performed to test the detection capabilities of the RMS. The removed assemblies were substituted with dummy assemblies to simulate concealment. The studies showed that the RMS design is suitable to monitor and detect the removal of even a single SNF assembly from the cask. [ABSTRACT FROM AUTHOR]

Understanding radiation effects on avalanche photodiodes (APDs) is important because they are used in several applications involving harsh radiation environments. APDs are used as photosensors in applications where speed and detection efficiency are critical. Proton irradiation experiments on a commercial off-the-shelf APD demonstrated that the irradiation flux and applied reverse bias have a strong influence on the severity of radiation effects. This is measured using the ion beam induced charge (IBIC) technique in which charge collection efficiency (CCE) describes the signal response from a device. CCE can degrade substantially due to radiation damage, but recent measurements show that certain combinations of irradiation flux and reverse bias can lead to increases in CCE up to 186% ± 24% for irradiations with 2 MeV protons at a fluence of $6.4\times 10^{11}$ cm−2. This defect-enhanced charge multiplication (DECM) only appeared when the reverse bias during irradiation ranged from 170 to 1830 V out of a maximum operating bias of 2000 V and the proton flux ranged from $9.8\times 10^{7}$ to $3.4\times 10^{9}$ cm−2s−1. Values outside of either range led to losses in CCE. It is expected that DECM should be encountered in other devices, especially those with sufficiently high electric fields to cause impact ionization. [ABSTRACT FROM AUTHOR]

Human Reliability Program (HRP) is a set of procedures, protocols, and the corresponding performance of activities to support and sustain the secure and safe operation of a facility. Our review indicated that HRP is referred to by various names, such as, Personnel Reliability Program (PRP), Trustworthiness Program, and Fitness for Duty Program, with PRP being the most common amongst them. A brief description of the evolution of HRP and its components for Finland, Japan, Jordan, Nigeria, Russia, US, and UK is provided. Details of the Trustworthiness Program by the International Atomic Energy Agency are also provided. Most of the HRPs have similar components and implementation steps with few variations, and they all have identical objectives. Even though a structured HRP may not exist in Indian civil nuclear energy facilities, an account of various HRP components practiced in India is discussed. The HRP review provided should help security practitioners and academics. [ABSTRACT FROM AUTHOR]

When the foundation of a method is simulated data, it is paramount that the method is validated with data from physical samples when possible. This study presents the results of validating a recently developed nuclear forensics methodology with a new low-burnup plutonium sample, chemically separated from low-enriched uranium irradiated in thermal neutron flux. The nuclear forensics methodology uses machine learning models to discriminate the reactor type of origin, fuel burnup, and time since irradiation (TSI) of chemically separated plutonium samples. The machine learning models use intra-elemental isotope ratios of cesium, samarium, europium, and plutonium as features; the isotopic ratio data for training the models were generated through fuel burnup simulations of various nuclear reactor types. The methodology predicted the reactor type and fuel burnup of the plutonium sample successfully. Initial difficulties to predict the TSI were resolved with the inclusion of a new intra-elemental isotope ratio of cerium. [ABSTRACT FROM AUTHOR]

The Monte Carlo N-Particle (MCNP6) radiation transport code is widely used to perform material transmutation and depletion calculations using the embedded module CINDER90. CINDER90 is capable of obtaining fission product and transuranic nuclide concentrations with a high level of accuracy in irradiated nuclear fuel. This information is very useful for many nuclear applications including reactor design and analysis, nuclear safeguards, nuclear security, and nuclear forensics, to name a few. However, at present the MCNP6 code does not estimate the overall statistical uncertainty in the nuclide concentrations reported at the end of a depletion calculation. We report our approach using a random sampling method to estimate stochastic uncertainty in fission product nuclide concentration using various parameters reported in MCNP6 output and how these uncertainties are affected by the calculation parameters. [ABSTRACT FROM AUTHOR]

There is a shortage of reliable information on the highly enriched uranium (HEU) stockpile, production capabilities, and natural uranium reserves of the Democratic People's Republic of Korea (DPRK or North Korea). It is, however, possible to estimate DPRK's nuclear material reserves using the data in the open literature and considering various scenarios. Based on our literature survey and analysis we are projecting DPRK's natural uranium reserves and their production capabilities of weapons-grade HEU. We also report the uncertainties associated with DPRK's uranium enrichment capabilities due to the differences in estimates provided in the literature. Our analysis shows that given the range of the estimates of DPRK's natural uranium ore reserves, its nuclear weapons program is unlikely to be constrained by uranium resources, provided they have the required mining and milling capacities. [ABSTRACT FROM AUTHOR]

As part of an experimental validation of a nuclear forensics methodology for Pu source reactor-type discrimination, destructive analysis has been performed on two irradiated UO2 pellets with different irradiation histories. Analysis has focused on measuring key Sm fission product isotope ratios used in a previously published maximum likelihood methodology to determine the most likely irradiation history of the pellets. A total of 21 Sm isotope ratios were measured within the irradiated pellets, and generally agreed within 20% of the irradiations as simulated using the Monte Carlo Radiation Transport and material depletion code, MCNP6. Results indicate the chosen approach can accurately measure the isotope ratios within 5% experimental error. [ABSTRACT FROM AUTHOR]

Under normal operating conditions, a pyroprocessing facility removes highly radioactive and nonradioactive fission product waste from used nuclear reactor fuel to recycle the remaining uranium (U), plutonium (Pu), and other actinides contained in it. The products from this facility are separate ingots of U and mixed transuranic elements (TRUs)-uranium (TRU-U). Uranium in both ingots will be depleted U with 235U enrichment less than 1%. The TRU-U ingot will contain neptunium, Pu, americium (Am), and curium (Cm) mixed with U with an approximate TRU:U ratio of 1:1. Four scenarios of nuclear material diversion by potential misuse of the pyroprocessing facility operations are analyzed and compared with the scenario of normal operating condition when the electrowinning process or the TRU-U ingot manufacturing process is misused. These diversion scenario analyses are carried out to understand the proliferation potential and to recommend safeguards measures. The four scenarios of nuclear material diversion analyzed are (1) 50 g Pu, (2) 100 g Pu, (3) 200 g Pu, and (4) all Pu, i.e., 452 g in the 1-kg TRU-U ingot. Plutonium cannot be diverted by itself because other TRUs (Am and Cm) will be simultaneously extracted with Pu. This is because the reduction potentials of those actinides are not distinguishably different from that of Pu on a liquid cadmium cathode of the electrowinning step of the pyroprocess. Hence, in addition to Pu, simultaneous diversion of respective amounts of Am and Cm for the four diversion scenarios are considered. The diversion scenario analysis also considered the concealment of Pu and Cm removal from the TRU-U ingot by adding an equivalent amount of 252Cf to replenish the neutron source emissions. These five scenarios (four nuclear material diversion scenarios and one normal operation scenario) are modeled and simulated using the Monte Carlo N-Particle (MCNP6) radiation transport computer code by incorporating the model of a NaI gamma radiation detection system. The results show that the presence and absence of Pu in the TRU-U ingot can be confirmed by the NaI gamma radiation detection system. However, identifying the presence of U in the TRU-U ingot is difficult using the NaI gamma radiation detection system due to interference from TRU gamma radiation. To identify the U presence in the TRU-U ingot, an application of nuclear magnetic resonance (NMR) is studied. The NMR technology employs a numerical calculation approach based on density functional theory (DFT) simulation. The DFT calculation results show that the detection of U in a pyroprocess is feasible by NMR technology. In addition, these four nuclear material diversion scenarios are analyzed through MCNP6 simulations by incorporating the model of a coincidence neutron detection system. To conceal the nuclear material diversion, the simulations are performed by replacing the diverted Pu and Cm by an appropriate mass of 252Cf neutron source that is equivalent to the neutron source strengths of the diverted mass. Simulation results show that this concealment (misuse) results in a deceived Pu mass estimate in the TRU-U ingot if the Pu-to-244Cm-ratio method (proposed method in the literature) is used. [ABSTRACT FROM AUTHOR]