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DOE OSTI2018

Implementation of Microcalorimeter Array Technology for Safeguards of Nuclear Material

Safeguards of nuclear materials depends on both destructive and nondestructive assay (DA and NDA respectively). Ultra-high resolution microcalorimeter gamma spectroscopy has the potential to substantially reduce the performance gap between NDA and DA methods in determination of plutonium isotopic composition. This paper details the setup of a cryostat and microwave readout system for microcalorimeter gamma spectroscopy, the functionality of which has been successfully demonstrated.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2015

Design and performance of A 3 He-free coincidence counter based on parallel plate boron-lined proportional technology

Thermal neutron counters utilized and developed for deployment as non-destructive assay (NDA) instruments in the field of nuclear safeguards traditionally rely on 3He-based proportional counting systems. 3 He-based proportional counters have provided core NDA detection capabilities for several decades and have proven to be extremely reliable with range of features highly desirable for nuclear facility deployment. Facing the current depletion of 3He gas supply and the continuing uncertainty of options for future resupply, a search for detection technologies that could provide feasible short-term alternative to 3 He gas was initiated worldwide. As part of this effort, Los Alamos National Laboratory (LANL) designed and built a 3 He-free full scale thermal neutron coincidence counter based on boron-lined proportional technology. The boronlined technology was selected in a comprehensive inter-comparison exercise based on its favorable performance against safeguards specific parameters. This paper provides an overview of the design and initial performance evaluation of the prototype High Level Neutron counter – Boron (HLNB). The initial results suggest that current HLNB design is capable to provide ~80% performance of a selected reference 3 He-based coincidence counter (High Level Neutron Coincidence Counter, HLNCC). Similar samples are expected to be measurable in both systems, however, slightly longer measurement times may be anticipated for large samples in HLNB. The initial evaluation helped to identify potential for further performance improvements via additional tailoring of boron-layer thickness.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2016

PWR and BWR spent fuel assembly gamma spectra measurements

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137 Cs, 154 Eu, and 134 Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

An unattended verification station for UF 6 cylinders: Field trial findings

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235 U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs to the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. In conclusion, analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that material diversion has not occurred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Validation and demonstration of the AEFC as a practical safeguards tool for inventory verification

The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) instrument designed to determine the residual fissile mass in irradiated fuel assemblies for safeguards verification purposes. This is done by actively interrogating an assembly with a neutron source and measuring the total (Singles) and correlated (Doubles) neutron count rates resulting from induced fissions in the irradiated nuclear fuel and relating those rates to the residual fissile mass using calibration curves. Comprehensive NDA measurements of the irradiated fuel inventory at Israeli Research Reactor 1 (IRR-1) were taken with the AEFC to validate a set of previously developed calibration curves. During the campaign, measurements were acquired of 32 standard fuel assemblies and three control assemblies in just nine days. This is a significant majority of the research reactor's irradiated fuel inventory and the largest data set gathered by the AEFC to date. Many of the fuel assemblies measured during the campaign had much shorter cooling times than those assemblies previously measured with the instrument. Calibration curves developed from previous AEFC deployments were used to determine the residual 235 U mass in the measured standard fuel assemblies. The results of the campaign demonstrated that the AEFC can be used to estimate the 235 U mass remaining in a large number of irradiated fuel assemblies with 1%–5% uncertainty in a reasonable amount of time despite operating in a high dose environment.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2018

A Weighted Least-Squares Transport Equation Compatible with Source Iteration and Voids

We present that second-order forms of the transport equation allow the use of continuous finite elements (CFEMs). This can be desired in multiphysics calculations where other physics require CFEM discretizations. Second-order transport operators are generally self-adjoint, yielding symmetric positive-definite (SPD) matrices, which allow the use of efficient linear algebra solvers with an enormous advantage in memory usage. Least-squares (LS) forms of the transport equation can circumvent the void problems of other second-order forms but are almost always nonconservative. Additionally, the standard LS form is not compatible with discrete ordinates method (SN) iterative solution techniques such as source iteration. A new form of the LS transport equation has recently been developed that is compatible with voids and standard SN iterative solution techniques. Performing nonlinear diffusion acceleration (NDA) using an independently differenced low-order equation enforces conservation for the whole system and makes this equation suitable for reactor physics calculations. In this context, “independent” means that both the transport and low-order solutions converge to the same scalar flux and current as the spatial mesh is refined, but for a given mesh, the solutions are not necessarily equal. In this paper we show that introducing a weight function into this LS equation improves issues with causality and can render our equation equal to the self-adjoint angular flux (SAAF) equation. Causality is a principle of the transport equation that states that information travels only downstream along characteristics. This principle can be violated numerically. We show how to limit the weight function in voids and demonstrate the effect of this limit on accuracy. Using the C5G7 benchmark, we compare our method to the SAAF formulation with a void treatment (SAAFτ) that is not self-adjoint and has a nonsymmetric coefficient matrix. Finallyl, we show that the weighted LS equation with NDA gives acceptable accuracy relative to the SAAFτ equation while maintaining a SPD system matrix.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

Interpretation of nasal swab measurements following suspected releases of actinide aerosols

For radionuclides such as plutonium and americium, detection of removable activity in the nose (i.e., nasal swab measurements) are frequently used to determine whether follow-up bioassay measurements are warranted following a potential intake. For this paper, the authors analyzed 429 nasal swab measurements taken following incidents or suspicious circumstances (such as an air monitor alarming) at Los Alamos National Laboratory (LANL) for which the dose was later evaluated using in vitro bioassay. Nasal swab measurements were found to be very poor predictors of dose and should not be used as such in the field. However, nasal swab measurements can be indicative of whether a reliably detectable committed effective dose (CED) occurred. About 14% of nasal swab measurements between 1.25 and 16.7 Bq corresponded to CEDs greater than 1 mSv, so in general, positive nasal swabs always indicate that follow-up bioassay should be performed (positive nasal swabs less than 1.25 Bq are considered separately). This probability increased significantly for nasal swabs greater than 16.7 Bq. Only about 3% of nasal swabs with no detectable activity (NDA) corresponded to reliably detectable CEDs. As a result, a nasal swab with NDA is therefore necessary, but not sufficient, to negate the need for a follow-up bioassay if it was collected following other workplace indicators of a potential intake.

238Pu↗

DOE OSTI2015

Neutron spectrometry for UF 6 enrichment verification in storage cylinders

Verification of declared UF 6 enrichment and mass in storage cylinders is of great interest in nuclear material nonproliferation. Nondestructive assay (NDA) techniques are commonly used for safeguards inspections to ensure accountancy of declared nuclear materials. Common NDA techniques used include gamma-ray spectrometry and both passive and active neutron measurements. In the present study, neutron spectrometry was investigated for verification of UF 6 enrichment in 30B storage cylinders based on an unattended and passive measurement approach. MCNP5 and Geant4 simulated neutron spectra, for selected UF 6 enrichments and filling profiles, were used in the investigation. The simulated neutron spectra were analyzed using principal component analysis (PCA). The PCA technique is a well-established technique and has a wide area of application including feature analysis, outlier detection, and gamma-ray spectral analysis. Results obtained demonstrate that neutron spectrometry supported by spectral feature analysis has potential for assaying UF 6 enrichment in storage cylinders. The results from the present study also showed that difficulties associated with the UF 6 filling profile and observed in other unattended passive neutron measurements can possibly be overcome using the approach presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

ARIES Oxide Production Program Monthly Report – April 2014

A summary of the accomplishments, milestones, financial summary, project performance and issues facing the Oxide Production Program for the end of April 2014 is presented in this Executive Summary. PF-4 and ARIES operations remain paused in order to resolve issues and perform an extent-of-condition review for Criticality Safety and Conduct of Operations. On April 28th, Jeff Yarbrough issued a memorandum updating status on the “pause” of PF-4 programmatic operations. He reported that 150 of the procedures that control programmatic operations in PF-4 have successfully resumed. These 150 procedures represent about ½ of the total that will enable us to meet FY14 programmatic commitments. No formal letter of direction will be provided to LANS via the NNSA Field Office contracting officer to address the direction relative to shipments, analytical chemistry sampling and guidance for financial adjustments in FY14 and FY15. LANS will provide a revised plan based on the funding profile included in the President’s Budget Request for the remainder of FY14 and FY15 to include revised FY14 milestones and proposed FY15 milestones. The ARIES Pit Cutter project met the NNSA milestone (Complete Design of Simple Pit Disassembly Equipment [ARIES Pit Cutter] – Due 8/15/14) by completing the fabrication design on 4/3/14, well in advance of the 8/15/14 due date. The project team has started procurement of long-lead items in preparation of the prototype fabrication and cold testing. The fabrication is scheduled to complete in FY14 and the cold testing will begin in FY14 and complete in early FY15. LANL and PNNL met on April 14 to discuss the path forward for PNNL criticality support to the program. The consensus was that PNNL resources can be effectively applied to the generation of detailed CSEs to support the program independent of the current ESS strategy. FY15 program planning is underway. The team is working on planning/drafting the work packages and developing the schedule under the assumption that operations will begin October 1, 2014. A planning schedule has been finalized and the team is meeting regularly to discuss details at the work package and WBS levels. The FY15 PMP is also in the process of being updated. The goal is to have all the drafts reviewed/approved and finalized prior to the start of the FY. MQ personnel have completed ADPSM Management Assessment: Evaluate Quality Roles and Responsibility Agreements between ADPSM and LANL Technical Support Organizations (PFITS 2011-1943). Two findings and four opportunities for improvement were reported. The robotic lathe controller upgrade has entered the construction phase of the project. The controller cabling IWD pre-job meeting was completed. The initial robotic lathe functional test was completed. The existing controller power supply disconnect is scheduled for early May. The NDA Cage project team received notification on 3/28/14 to restart the NDA cage project. Since then the team have been busy with activities that include assigning a planner, getting the construction work package ready for construction and getting the USQ completed. The electrical services installation for the new welding skid for the Inner Can Welder was completed. A gas bottle rack, used to support welding operations, will be installed in early May. The Outer Can Welder design is complete and the IWD is in development to execute construction. Installation materials have been ordered. Work continues on the MD-2 preliminary design. The project engineer is performing a technical evaluation of the Project Management Plan that was submitted by the design agent in response to the LANL Request for Proposal. If all requirements have been captured, LANL procurement will develop a contract and give the design agent notice to proceed on the preliminary design. Estimated completion is scheduled for September 2014. The design for the TA-55 Bldg. 314 warehouse was issued at 90% complete and has begun the last review process. A constructability review is scheduled for early May. The Project Manager has been working with the LANL estimating group to get the class 3 estimate for review. The Oxide Production Cage Upgrade project has kicked-off efforts on the 30% design. Requirements documentation has been drafted. The floor space was walked-down with other project managers and stakeholders and the boundaries were agreed upon.

ARIES↗

DOE OSTI2015

Fundamentals of Nondestructive Assay [Slides]

NDA measurements play an important role in managing nuclear material through their ability to detect, identify, locate, and often quantify the composition and/or mass of material within a relatively short amount of time. The three main passive NDA measurement techniques that are utilized in managing nuclear material are (1) Gamma-ray spectroscopy (composition); (2) Neutron Multiplicity counting (fission rate of Pu items); and (3) Calorimetry (heat production from decay of Pu bearing items). Through combination of Gamma-ray spectroscopy and either neutron multiplicity counting or calorimetry, mass of Pu bearing items can be measured.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2015

Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

High-dose neutron detector development

The development of advanced sustainable nuclear fuel cycles relying on used nuclear fuel is one of the key programs pursued by the DOE Office of Nuclear Energy to minimize waste generation, limit proliferation risk and maximize energy production using nuclear energy. Safeguarding of advanced nuclear fuel cycles is essential to ensure the safety and security of the nuclear material. Current non-destructive assay (NDA) systems typically employ fission chambers or 3He-based tubes for the measurement of used fuel. Fission chambers are capable of withstanding the high gamma-ray backgrounds; however, they provide very low detection efficiency on the order of 0.01%. To benefit from the additional information provided by correlated neutron counting [1] higher detection efficiencies are required. 3He-based designs allow for higher detection efficiencies; however, at the expense of slow signal rise time characteristics and higher sensitivity to the gamma-ray backgrounds. It is therefore desirable to evaluate and develop technologies with potential to exceed performance parameters of standard fission chamber-based or 3He-based detection systems currently used in the NDA instrumentation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Year End Progress Report on Rattlesnake Improvements

Rattlesnake is a MOOSE-based radiation transport application developed at INL to support modern multi-physics simulations. At the beginning of the last year, Rattlesnake was able to perform steady-state, transient and eigenvalue calculations for the multigroup radiation transport equations. Various discretization schemes, including continuous finite element method (FEM) with discrete ordinates method (SN) and spherical harmonics expansion method (PN) for the self-adjoint angular flux (SAAF) formulation, continuous FEM (CFEM) with SN for the least square (LS) formulation, diffusion approximation with CFEM and discontinuous FEM (DFEM), have been implemented. A separate toolkit, YAKXS, for multigroup cross section management was developed to support Rattlesnake calculations with feedback both from changes in the field variables, such as fuel temperature, coolant density, and etc., and in isotope inventory. The framework for doing nonlinear diffusion acceleration (NDA) within Rattlesnake has been set up, and both NDA calculations with SAAF-SN-CFEM scheme and Monte Carlo with OpenMC have been performed. It was also used for coupling BISON and RELAP-7 for the full-core multiphysics simulations. Within the last fiscal year, significant improvements have been made in Rattlesnake. Rattlesnake development was migrated into our internal GITLAB development environment at the end of year 2014. Since then total 369 merge requests has been accepted into Rattlesnake. It is noted that the MOOSE framework that Rattlesnake is based on is under continuous developments. Improvements made in MOOSE can improve the Rattlesnake. It is acknowledged that MOOSE developers spent efforts on patching Rattlesnake for the improvements made on the framework side. This report will not cover the code restructuring for better readability and modularity and documentation improvements, which we have spent tremendous effort on. It only details some of improvements in the following sections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

A kinematically complete, interdisciplinary, and co-institutional measurement of the 19 F(α,n) cross section for nuclear safeguards science

Alpha particles emitted from the decay of uranium in a UF 6 matrix can interact with fluorine and generate neutrons via the 19 F(α,n) 22 Na reaction. These neutrons can be used to determine the uranium content in a UF 6 storage cylinder. The accuracy of this self-interrogating, non-destructive assay (NDA) technique is, however, limited by the uncertainty of the 19 F(α,n) 22 Na cross section. We have performed complementary measurements of the 19 F(α,n) 22 Na reaction with both 4 He and 19 F beams to improve the precision of the 19 F(α,n) 22 Na cross section over the alpha energy range that encompasses common actinide alpha decay needed for NDA studies. We have determined an absolute cross section for the 19 F(α,n) 22 Na reaction to an average precision of 7.6% over the alpha energy range of 3.9 – 6.7 MeV. We utilized this cross section in a simulation of a 100 g spherical UF 6 assembly and obtained a change in neutron emission rate values of approximately 10-12%, and a significant (factor of 3.6) decrease in the neutron emission rate uncertainty (from 50-51% to 13-14%), compared to simulations using the old cross section. Our new absolute cross section enables improved interpretations of NDAs of containers of arbitrary size and configuration.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2016

Analysis of Selected Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened, and the inner package, removed and placed in a radiological hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear intact and to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 had been opened at an undetermined time in the past. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the nondestructive analysis (NDA) results, a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples, it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion. The materials contained in Capsules 2 and 5 have been examined. There appears to be a relatively uniform distribution of Kr and Rb throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within Capsule 5 showed ~1 at. % lead. The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation / sintering process that occurred when the samples were made. The capsule walls were also examined and showed extensive corrosion throughout. Elemental mapping of the capsule material appeared consistent with carbon steel, while the weld material appeared consistent with a stainless steel. The interior surface of the capsule appeared to have a layer of material containing Al, Si, and Ca similar to the 5A molecular sieve. Analysis for Rb within the corrosion sites was inconclusive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly to confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Progress, Performance, and Prospects of Ultra-High Resolution Microcalorimeter Spectrometers

In 2005 the LANL/NIST team used a single high-resolution microcalorimeter detector to measure the gamma-ray spectrum of a plutonium sample. After more than a decade of research and development on this topic, both the technology and our general understanding of its capabilities have advanced greatly, such that a progress review is now timely. We examine the scenario of a large-scale reprocessing plant and conclude that current non-destructive analysis (NDA) methods are inadequate to safeguard such a facility to the desired levels, leading to undesirable dependence on massspectrometry (MS) destructive analysis (DA). The development of microcalorimeter detectors is intended to close the performance gap between NDA and DA methods to address the needs of nuclear facilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗