Sunil S. Chirayath

Detecting nuclear materials smuggling: using radiography to improve container inspection policies

This paper proposes a layered container inspection system for detecting illicit nuclear materials using radiography information. We argue that the current inspection system, relying heavily on the Automated Targeting System (ATS) and passive radiation detectors, is inherently incapable of reliably detecting shielded radioactive materials, especially highly enriched uranium (HEU). This motivates the development of a new inspection system, which is designed to address a fundamental flaw of the ATS-based system, allowing for improved defense against sophisticated adversaries. In the proposed inspection system, all cargo containers go through x-ray imaging equipment first. From the x-ray image, a hardness measure of the container is computed. This hardness measure characterizes how likely it is that shielded HEU, if it does exist in the container, will not be detected in a subsequent passive detection step. Depending on the value of the hardness, the lower-hardness containers are sent to passive detection and the high-hardness containers are sent directly to active detection. This paper explores the trade-off between the detection probability of the new inspection system and the expected sojourn time a container spends in the system. The solution details and decision-making tools for using such a system are provided. Comparisons are made between the proposed system and the current ATS-based nuclear inspection system.

Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast and Thermal Reactors

The growing concern about nuclear terrorism threats has enhanced the need to develop fast and accurate nuclear forensics analysis techniques for nuclear material source attribution and to create a credible nuclear deterrence. Plutonium produced as a by-product in nuclear reactor fuel, especially in fuel discharged at low burn-up (1 to 2 MWd/kg), is potentially weapons usable material. In the event of plutonium interdiction from a smuggling act, its origin has to be established through nuclear forensics attribution methods before any response is initiated against this malicious act. The characteristics of separated plutonium from discharged reactor fuel and the associated fission product traces depend on factors such as the reactor type (thermal or fast reactor), fuel burn-up, irradiation history, and the chemical process used to separate plutonium. A new methodology of using trace fission product to plutonium ratios for nuclear forensics attribution of plutonium to the type of reactor used for its production is presented along with results obtained for case studies of a fast neutron spectrum breeder reactor and a thermal neutron spectrum reactor using open literature design information of these two types of nuclear reactors.

Experimental and Computational Forensics Characterization of Weapons-Grade Plutonium Produced in a Fast Reactor Neutron Environment

A terrorist attack using an improvised nuclear device is one of the most serious dangers facing the United States. The work presented here is part of an effort to improve nuclear deterrence by developing a methodology to attribute weapons-grade plutonium to a source reactor by measuring the intrinsic physical characteristics of the interdicted plutonium. In order to demonstrate the developed methodology, plutonium samples were produced from depleted uranium dioxide (DUO2) surrogates irradiated in a fast-neutron environment. In order to replicate the neutron flux in a fast-neutron-spectrum reactor and obtain experimental samples emulating weapons-grade plutonium produced in the blanket of a fast breeder reactor, DUO2 samples were placed in a gadolinium sheath and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Previous computational work on this topic identified several fission products that could be used to distinguish between reactor types (fast and thermal reactors), specifically: 137Cs, 134Cs, 154Eu, 125Sb, 144Ce, 85Rb, 147Pm, and 150Sm along with the plutonium isotopes. Simulations of the fast neutron irradiation of the DUO2 fuel surrogates in the HFIR were carried out using the Monte Carlo radiation transport code MCNPX 2.7. Comparisons of the predicted values of plutonium and fission product concentrations to destructive and nondestructive assay measurements of neutron-irradiated DUO2 surrogates are presented here. The agreement between the predictions and gamma spectroscopic measurements in general were within 10% for 134Cs, 137Cs, 154Eu, and 144Ce. Additional experimental results (mass spectroscopy) agreed to within 5% for the following isotopes: 85Rb, 147Pm, 150Sm, 154Eu, 148Nd, 144Ce, and 239Pu. Two indicator isotopes previously suggested to differentiate between the reactor types were ruled out for use in the attribution methodology; 125Sb was ruled out due to the difficulty in accurately predicting its concentration, and 242Pu was ruled out because of its low content in weapons-grade plutonium.

Nuclear Forensics Methodology for Reactor-Type Attribution of Chemically Separated Plutonium

Abstract 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 isotopes selected for the attribution methodology development were based upon the initial criteria of isotope production yield in fuel and half-life. Subsequently, intra-element isotope ratios were formed with the criterion that the ratio must have a functional dependence on at least one of the reactor parameters of interest. The developed methodology compares the values of reactor-dependent intra-element isotope ratios in the library developed to the same ratios of the interdicted sample. A maximum likelihood calculation methodology was utilized to perform the aforementioned multiple intra-element isotope ratio comparison to produce a single metric to depict the result of the comparison. The methodology can predict the reactor type, fuel burnup, and time since irradiation of the sample by selecting the array of reactor-dependent intra-element isotope ratios that provides the maximum likelihood value. The methodology was tested with intra-element ratios of pseudo interdicted sample data and found to be viable for source attribution.

Nuclear Security Risk Analysis: An Insider-Outsider Collusion Scenario

This study analyses the vulnerability of the physical protection system (PPS) deployed at a hypothetical facility. The PPS is designed to prevent and eliminate threats to nuclear materials and facilities. The analysis considers possible outsider and insider threats. A modified adversary sequence diagram (ASD) evaluates threat pathways to test an insider-outsider collusion case. The ASD also measures the probability of adversary interruption by demonstrating the methodology for a typical nuclear facility.

Fission product decontamination factors for plutonium separated by PUREX from low-burnup, fast-neutron irradiated depleted UO2.

Experimental investigations to determine fission product separation from actinides (U and Pu) while employing the Plutonium Uranium Recovery by Extraction (PUREX) process to purify plutonium produced in a fast neutron irradiated depleted uranium dioxide (DUO2) target were conducted. The sample was a DUO2 pellet (0.256wt% 235U) irradiated to a low-burnup (4.43±0.31GWd/tHM) that was PUREX processed 538 days after neutron irradiation. Decontamination factors (DF) for the elements U, Mo, Ru, Ce, Sm, Sr, Pm, Eu, Nd, Pd, and Cd were measured with mass spectroscopy in two experiments using 30vol% tri-n-butyl phosphate (TBP) in a kerosene diluent. The first experiment characterized Pu DFs for a single stage extraction and back-extraction, while the second experiment had multiple stages with the goal of achieving greater Pu recovery. The benchtop scale PUREX process had overall Pu recoveries of (83.4±9.5)% and (99.7±4.2)% for the first and second experiments, respectively.

Investigations on Detecting Potential Nuclear Material Diversion from a Pyroprocessing Facility

Abstract 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.

Uncertainty Quantification of Concrete Utilized in Dry Cask Storage

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 compositions and densities. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The model was developed for use in calculations with MCNP, with reference data from a UNF assembly source and geometry details based on the Holtec HI-STORM 100S UNF dry cask storage system, both provided by the Comanche Peak nuclear power plant. A number of cases have been developed and analyzed to compare the effects of variations in concrete compositions considering nominal densities, variations from nominal densities, fixed densities regardless of the specific composition, and variations in the decay source. The analysis confirmed that the parametric variability of concrete compositions is a major source of uncertainty in evaluations of dry cask dose rates. While precise results depend on the compositions compared, general trends can be identified. The largest fraction of the dose value in all cases, typically 70%, is due to gamma rays produced by the fission products. Density variation had a dominant effect on the dose rate. Composition variations, while density was held fixed, indicated that the specific composition data significantly impact the dose rates produced by neutrons and associated capture gamma rays. The impact due to composition variation on neutron dose rate was found to be on the order of 70% or higher for these test cases. The analysis indicates that uncertainties in concrete characteristics at the time of on-site pour procedures impact the actual shielding efficiency and, therefore, must be evaluated.

The Nuclear Security Science and Policy Institute at Texas A&M University

NSSPI is a multidisciplinary organization at Texas AM (2) educate the next generation of nuclear security and nuclear nonproliferation leaders; (3) analyze the interrelationships between policy and technology in the field of nuclear security; and (4) serve as a public resource for the reduction of nuclear threats. Since 2006, Texas AM (2) advise students on valuable research projects that contribute substantially to the nuclear nonproliferation, safeguards, and security arenas; and (3) engage with experts from several similar international academic and research institutes in activities and research that benefit Texas A&M students. NSSPI also helps international institutions develop their own programs in nuclear security and nonproliferation.

Assessment of the Fingerprinting Method for the Verification of Spent Fuel in MACSTOR KN-400 CANDU Spent-Fuel Dry Storage

Abstract Korea Hydro and Nuclear Power has built a new modular type of CANDU spent fuel bundle dry storage facility, MACSTOR KN-400, at the Wolsong reactor site in the Republic of Korea. Four CANDU reactors operate at the Wolsong site, and the MACSTOR KN-400 has the capacity to store up to 24 000 CANDU spent fuel bundles. The International Atomic Energy Agency safeguards regulations demand an effective method for spent-fuel re-verification at the MACSTOR KN-400 facility in the event of any loss of continuity of knowledge. A radiation signal-dependent spent-fuel re-verification design of the MACSTOR KN-400 is scrutinized through mathematical model development and Monte Carlo radiation transport simulations using the state-of-the-art computer code MCNP. Both gamma and neutron transport simulations for various spent fuel bundle diversion scenarios are carried out for the central and corner re-verification tube structures. The CANDU spent fuel bundles with a burnup of 7500 MWd/tonne U (burned at a specific power of 28.39 MW/tonne) and 10 years of cooling time are considered for the radiation source term. Results of the gamma transport simulations incorporating cadmium-zinc-telluride detectors inside the re-verification tube show that spent fuel bundles diverted from the inner locations of the storage basket cannot be detected by observing a gamma radiation signal change. Neutron transport simulations consisting of a 3He detector inside the re-verification tube show that certain spent fuel bundle diversions could be detected. However, inverse MCNP neutron transport simulations show that the possibility of detecting diversion of ~67% of spent fuel bundles stored in the basket region on the opposite side from the collimator of the re-verification tube is small, assuming a neutron detection counting time of 1 h per re-verification tube. It is also observed that the nondetection probability for most of the diversion scenarios considered is large. Nondetection probability here is defined as the probability of not detecting the diversion of spent fuel bundles from the baskets by observing radiation signal reduction from the removal of the bundles. Containment and surveillance methods are being employed for safeguards purposes at the facility, supplemented by periodic axial profile fingerprinting. However, since the nondetection probability is large for most scenarios, the facility should consider alternatives to this method in case loss of continuity of knowledge occurs.