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A recently published nuclear forensics methodology for source discrimination of separated weapons-grade plutonium utilizes intra-element isotope ratios and a maximum likelihood formulation to identify the most likely source reactor-type, fuel burnup and time since irradiation of unknown material. Sensitivity studies performed here on the effects of random measurement error and the uncertainty in intra-element isotope ratio values show that different intra-element isotope ratios have disproportionate contributions to the determination of the reactor parameters. The methodology is robust to individual errors in measured intra-element isotope ratio values and even more so for uniform systematic errors due to competing effects on the predictions from the selected intra-element isotope ratios suite. For a unique sample-model pair, simulation uncertainties of up to 28% are acceptable without impeding successful source-reactor discrimination. However, for a generic sample with multiple plausible sources within the reactor library, uncertainties of 7% or less may be required. The results confirm the critical role of accurate reactor core physics, fuel burnup simulations and experimental measurements in the proposed methodology where increased simulation uncertainty is found to significantly affect the capability to discriminate between the reactors in the library. Keywords: Nuclear forensics, Reactor-type discrimination, Maximum likelihood, Intra-element isotope ratios, Sensitivity analysis

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 capab...

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.

An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed. Keywords: Nuclear forensics, Reactor-type discrimination, Weapons-useable plutonium, Intra-element isotope ratios, Maximum likelihood

The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural UO2 fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately 20 μg of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural UO2 fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural UO2 fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%. Keywords: Neutron Irradiation, Nuclear Forensics, Weapons-grade Plutonium

The co-crystallization of hexavalent actinides with uranyl nitrate hexahydrate has been established as a viable approach to achieving group actinide separation in used nuclear fuel recycle, while avoiding the difficulties associated with the solvent-extraction process altogether. A preliminary neutronics safety analysis has been performed to alleviate criticality concerns associated with this group hexavalent actinide co-crystallization approach. The basic uranyl nitrate hexahydrate material compositions used are representative of used fuel found at commercial pressurized water reactors all around the U.S. This material is subsequently doped with various percentages of either typical reactor grade plutonium or with the total transuranic vector, replacing the uranium atoms in the crystal lattice structure. Infinite and effective neutron multiplication factors were calculated using the state-of-the-art Monte Carlo neutron transport code MCNP in a variety of plausible material configurations, assuming an acceptable safety limit for criticality to be an effective neutron multiplication factor less than 0.95. The results indicate that the crystal structures form a very subcritical intrinsic geometry even with six water molecules being incorporated into the crystal structure. The results of this research suggest that significant opportunity exists for the safe transfer of the group actinide separation approach from a laboratory to an operational scale.

A nuclear forensics methodology has been developed that is capable of source attribution of separated weapons-grade plutonium in case of an interdiction. The methodology utilizes plutonium and contaminant fission product isotopes within the separated plutonium sample to determine the characteristics (reactor parameters) of the interdicted material. The reactor parameters of interest include source reactor type, fuel irradiation burnup, and time since irradiation. The MCNPX-2.7 radiation transport code was used to model reactor cores and perform neutronics simulations to estimate the resulting isotopes of irradiated UO2 fuel. The simulation results were used to create a reactor-dependent library of irradiated fuel isotope ratio values as a function of fuel burnup and time since irradiation. Ratios of intra-element isotopes (fission product or actinide) are used as characteristics to determine a combination of reactor parameters of interest that could have produced the interdicted sample. The isotop...

A multi-physics computational methodology to analyze xenon induced power oscillations in a nuclear reactor is developed and presented. The methodology development takes into account both neutron transport and thermal hydraulics behaviors and their inter-dependent feedbacks in a nuclear power reactor as a function of fuel burn-up. The methodology uses the Monte Carlo N-Particle radiation transport computational code (MCNP6) along with its fuel transmutation module (CINDER90), a semi-analytic single channel analysis tool for thermal hydraulics and the SIGACE code suite for temperature dependent neutron interaction cross section processing. A Python script developed couples the physics codes for full automation. The accuracy of the multi-physics computational methodology developed was verified through a benchmark calculation for the published core parameters of a startup test performed at Yonggwang Power Reactor Unit No. 3. The power axial offset and xenon axial offset parameters were calculated for this benchmark case and used to quantify the oscillatory behavior observed, the results of this benchmark study is also presented here. The results showed that the developed methodology was able to capture the underlying phenomena governing xenon induced power oscillations in a nuclear reactor.

The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500 MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO2) fuel assemblies; 5 assemblies at 4.4 at% 235U enrichment and 64 assemblies at 4.95 at% 235U enrichment. The active fuel length is 200 cm and the core diameter is 194.55 cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240 cm resulting in an increased capacity factor of 95% at 530 MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.

We present a new stochastic computational tool, Multi-path Analysis of Physical Protection Systems (MAPPS) useful for Physical Protection System (PPS) effectiveness analysis. MAPPS tool can perform multi-path (adversary paths) analysis using an Adversary Sequence Diagram (ASD) and can determine the Most Vulnerable Path (MVP). The MVP determination by MAPPS uses the concept of Critical Detection Point (CDP). Suitability of MAPSS for PPS design effectiveness analysis is confirmed through an example. A comparison of MAPPS results with the traditional single path analysis (employing the Estimate of Adversary Sequence Interruption-EASI model) shows the advantages of using MAPPS. MAPPS correctly predicts the MVP and the distribution of probability of interruption (PI) values for various adversary strategies including collusion with facility insiders. MAPPS results show that a focusing only on mean of PI value distribution may not be appropriate instead, a analyzing the low probability tail of the PI distribution is important.

The objective of this paper is to analyze the uncertainty in radiation dose rate estimates outside of a used nuclear fuel (UNF) dry cask storage unit due to the parametric variability of concrete c...