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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Challenges and Recent Developments of Neutron Non-Destructive Assay with UO{sub 2}F{sub 2}-Bearing Materials - 17336

In the enrichment process of an operating Gaseous Diffusion Plant (GDP), such as the Portsmouth and Paducah facilities, UF6 gas will immediately hydrolyze forming the chemical compound UO{sub 2}F{sub 2} (Uranyl Fluoride). This mostly happens in Alumina (Al{sub 2}O{sub 3}) traps where the UF6 adsorbs to moistened Alumina pellets forming a UO{sub 2}F{sub 2} layer around the pellet. Similarly, NaF.UF{sub 6} absorption traps produce the complex Na{sub 2}UF{sub 8} which also hydrolyzes when exposed to moist air. UO{sub 2}F{sub 2} can also be formed in many other processes where UF6 gas comes into contact with moisture within structures and objects such as pipes, converters compressors and coolers. When the structures are refurbished, or designated for D and D, UO{sub 2}F{sub 2}-bearing objects are generated both in situ and ex situ requiring fissile (U-235) characterization. In addition, ex situ waste streams are generated when UO{sub 2}F{sub 2}-bearing materials are placed in drummed or boxed waste containers. Gamma spectroscopy Non-Destructive Assay (NDA) is commonly utilized to characterize UO{sub 2}F{sub 2}-bearing waste but generally has high associated Total Measurement Uncertainties (TMU's) and, in some cases, Data Quality Objectives (DQO's) cannot be met. In addition, high UO{sub 2}F{sub 2} mass loadings, such as hold-up materials, require gamma modeling when infinitely thick to the Uranium energy lines. A common neutron NDA technique, or method, is to exploit the neutrons generated from the F-19(a, n)Na-22 reaction which has an appreciable reaction cross-section. The isotopes of Uranium are a-emitters, and, in particular, U-234 has a significantly higher a-decay reaction cross-section leading to a significant relative neutron Specific Activity (nSA). Neutron NDA is appealing since, in some cases, it has potentially lower TMU's than gamma NDA and neutron detector slabs facilitate surrounding large, complex-shaped, objects. Characterizing UO{sub 2}F{sub 2}-bearing waste for fissile content using neutrons generated from U-234, requires the U-235/U-234 ratio as well as the enrichment as well as the corresponding nSA's. It is also noted that U-238 has a considerable, relative, spontaneous fission nSA and this is a higher concern for low-enrichments where the mass of U-238 dominates. Additional neutron interfering elements include cosmic-ray induced spallation within the objects and background from other, in situ, UO{sub 2}F{sub 2}-bearing objects. In this paper we present the challenges and recently developed hardware and algorithmic strategies for quantifying fissile mass both in situ and ex situ using neutron NDA that works in conjunction to gamma spectroscopy NDA, or process knowledge, for objects and drummed/boxed waste containing UO{sub 2}F{sub 2}-bearing materials leading to minimal TMU's and achievable DQO's. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Development of Non Destructive Assay Approach for Characterization of Spent Fuel Debris at Fukushima - 18543

Fukushima is considering options for decommissioning the damaged reactor. One of the technical challenges is the accurate characterization of fissile isotopes in the spent fuel debris. This is due to the fissile content, high gamma and neutron emission and uncertainties regarding the fuel burnup and unknown, heterogeneous mixture of materials that may be present. These requirements lead to significant challenges for potential measurement approaches with regard to the requirements for shielding and remote operation that result from plant and operational constraints. Mirion Technologies (MTKK) has been working to perform feasibility and concept design studies for a non-destructive assay (NDA) system that could be used to characterize the fissile content of retrieved samples of fuel debris. The main advantage of using an NDA technique is that the entire volume of the sample is interrogated so the method is not reliant upon assumptions of representativeness of samples retrieved and sent for laboratory analysis. The proposed technique comprises a combination of gamma spectroscopy and combined passive / active neutron interrogation with advanced data analysis algorithms in order to determine the maximum available data on the spent fuel properties and obtain a full radionuclide characterization of each sample. This includes quantification of individual fissile nuclides with validated uncertainties. In this paper we describe the NDA techniques and systems that are proposed, their benefits and limitations, and how they may be implemented into a full-scale system for fuel element debris characterization. We describe the studies that were performed based on experience of real systems and research supplemented with extensive computer modelling work to represent the expected fuel element debris properties. The work benefits from Mirion's extensive global physics and engineering expertise in designing, supplying and operating a wide range of NDA systems for diverse applications and in different challenging physical and operational environments. We describe how the advanced and innovative physics techniques that have resulted from Mirion's collective experience in these areas, can be deployed in an NDA system. The final approach that has been developed provides a means for determining the masses of the individual fissile isotopes with robust uncertainties and capable of implementation into an NDA system that meets the expected requirements for an operational plant. We focus on the techniques that have been developed and describe how they are of wider relevance and interest for projects that include challenges for characterisation of fuel residues. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Using Directional Diffusion Coefficients for Nonlinear Diffusion Acceleration of the First Order SN Equations in Near-Void Regions

The common definition of the diffusion coeffcient as the inverse of three times the transport cross section is not compat- ible with voids. Morel introduced a non-local tensor diffusion coeffcient that remains finite in voids[1]. It can be obtained by solving an auxiliary transport problem without scattering or fission. Larsen and Trahan successfully applied this diffusion coeffcient for enhancing the accuracy of diffusion solutions of very high temperature reactor (VHTR) problems that feature large, optically thin channels in the z-direction [2]. It is demonstrated that a significant reduction of error can be achieved in particular in the optically thin region. Along the same line of thought, non-local diffusion tensors are applied modeling the TREAT reactor confirming the findings of Larsen and Trahan [3]. Previous work of the authors have introduced a flexible Nonlinear-Diffusion Acceleration (NDA) method for the first order S N equations discretized with the discontinuous finite element method (DFEM), [4], [5], [6]. This NDA method uses a scalar diffusion coeffcient in the low-order system that is obtained as the flux weighted average of the inverse transport cross section. Hence, it su?ers from very large and potentially unbounded diffusion coeffcients in the low order problem. However, it was noted that the choice of the diffusion coeffcient does not influence consistency of the method at convergence and hence the di?usion coeffcient is essentially a free parameter. The choice of the di?usion coeffcient does, however, affect the convergence behavior of the nonlinear di?usion iterations. Within this work we use Morel’s non-local di?usion coef- ficient in the aforementioned NDA formulation in lieu of the flux weighted inverse of three times the transport cross section. The goal of this paper is to demonstrate that significant en- hancement of the spectral properties of NDA can be achieved in near void regions. For testing the spectral properties of the NDA with non-local diffusion coeffcients, the periodical horizontal interface problem is used [7]. This problem consists of alternating stripes of optically thin and thick materials both of which feature scattering ratios close to unity.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Verification of Special Nuclear Material (SNM) using Active Well Coincidence Counter (AWCC) and a High Purity Germanium (HPGe) Detector

SNM verification can be accomplished via two methods i.e. destructive analysis (DA) or nondestructive analysis (NDA). The DA modes are much more accurate compared to NDA modes; nevertheless, the sample's physical integrity is destroyed in the process. Techniques such as ICP-MS, ID-ICP-MS, TI-MS are few examples of DA, where a bulk material is ground and mixed with a solution (some form of acid) to acquire a homogenous sample. In contrast, NDA modes utilize signals emitted by the sample such as gamma, neutrons, light, or heat to analyze samples under investigation. HPGe detectors are some of the most frequently used gamma detectors to assay SNM. However, in a high activity scenario it provides a large uncertainty and sometimes becomes unfeasible to measure low energy gamma rays. Meanwhile, neutron detectors are not affected by the gamma background, therefore they can be used in addition to the gamma detectors to verify total fissile mass ({sup 235}U + {sup 239}Pu) of SNM. The AWCC is one of the thermal neutron NDA systems developed at the Los Alamos National Laboratory (LANL). It consists of 42 {sup 3}He detectors to measure gross neutrons (singles), coincidence neutrons (doubles), higher order multiplicity counting (triples, quadruples, etc.). It can effectively operate in both active and passive neutron interrogation modes. It can utilize AmLi, AmBe, or {sup 252}Cf active interrogation neutron sources. It is used to verify declared fissile mass ({sup 235}U or {sup 239}Pu) present in SNM. Objectives: Calibrate HPGe and AWCC for lightly irradiated (with enough {sup 137}Cs gamma background) ∼93% enriched HEU fuels; Verify fissile mass ({sup 235}U +{sup 239}Pu) content using AWCC and HPGe; Model and benchmark AWCC in MCNP6. Neutron coincidence and multiplicity counting: Two or more neutrons are coincident if they are detected by the system within the specified gate window; In AWCC: Gate window is 128 μsec; Induced fission in fissile material such as {sup 235}U and {sup 239}Pu - emission of time correlated neutrons - Singles, Doubles, Triples, or Quadruples; No induced fission in non-fissile materials- no correlated neutrons - only singles. SNM can be verified with high accuracy and precision using gamma and neutron NDA instruments such as HPGe and AWCC. Interaction with thermal neutrons. Non-fissile atoms - do not emit correlated neutrons. Fissile atoms - emit correlated neutrons during induced fission events. The time correlated neutrons can be detected by using coincidence or multiplicity counters such as AWCC. Likewise, {sup 235}U produces 186 keV gamma peak that can be measured by a HPGe gamma detector. Counts under 186 keV gamma peak provide information about {sup 235}U content. Combining AWCC and HPGe results can provide better understanding of special nuclear material under investigation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Using Directional Diffusion Coefficients for Nonlinear Diffusion Acceleration of the First Order S{sub N} Equations in Near-Void Regions

The common definition of the diffusion coefficient as the inverse of three times the transport cross section is not compatible with voids. Morel introduced a non-local tensor diffusion coefficient that remains finite in voids. It can be obtained by solving an auxiliary transport problem without scattering or fission. Larsen and Trahan successfully applied this diffusion coefficient for enhancing the accuracy of diffusion solutions of very high temperature reactor (VHTR) problems that feature large, optically thin channels in the z-direction. It was demonstrated that a significant reduction of error can be achieved in particular in the optically thin region. Along the same line of thought, non-local diffusion tensors have been applied to modeling the TREAT reactor confirming the findings of Larsen and Trahan. Previous work of the authors have introduced a flexible Nonlinear-Diffusion Acceleration (NDA) method for the first order S{sub N} equations discretized with the discontinuous finite element method (DFEM). This NDA method uses a scalar diffusion coefficient in the low-order system that is obtained as the flux weighted average of the inverse transport cross section. Hence, it suffers from very large and potentially unbounded diffusion coefficients in the low order problem. However, it was noted that the choice of the diffusion coefficient does not influence consistency of the method at convergence and hence the diffusion coefficient is essentially a free parameter. The choice of the diffusion coefficient does, however, affect the convergence behavior of the nonlinear diffusion iterations. Within this work we use Morel's non-local diffusion coefficient in the aforementioned NDA formulation in lieu of the flux weighted inverse of three times the transport cross section. The goal of this paper is to demonstrate that significant enhancement of the spectral properties of NDA can be achieved in near void regions. For testing the spectral properties of the NDA with non-local diffusion coefficients, the periodical horizontal interface problem is used. This problem consists of alternating stripes of optically thin and thick materials both of which feature scattering ratios close to unity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Nonlinear Diffusion Acceleration of the Least-Squares Transport Equation in Geometries with Voids

In this paper we show the extension of the Nonlinear-Diffusion Acceleration (NDA) to geometries containing small voids using a weighted least-squares (WLS) high order equation. Even though the WLS equation is well defined in voids, the low-order drift diffusion equation was not defined in materials with a zero cross section. This paper derives the necessary modifications to the NDA algorithm. We show that a small change to the NDA closure term and a non-local definition of the diffusion coefficient solve the problems for voids regions. These changes do not affect the algorithm for optical thick material regions, while making the algorithm well defined in optically thin ones. We use a Fourier analysis to perform an iterative analysis to confirm that the modifications result in a stable and efficient algorithm. Numerical results of our method will be presented in the second part of the paper. We test this formulation with a small, one-dimensional test problem. Additionally we present results for a modified version of the C5G7 benchmark containing voids as a more complex, reactor like problem. We compared our results to PDT, Texas A&M’s transport code, utilizing a first order discontinuous formulation as reference and the self-adjoint angular flux equation with void treatment (SAAFτ ), a different second order form. The results indicate, that the NDA WLS performed comparably or slightly worse then the asymmetric SAAFτ , while maintaining a symmetric discretization matrix.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Viability Study for an Unattended UF 6 Cylinder Verification Station: Phase I Final Report

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) that could provide automated, independent verification of the declared relative enrichment, 235 U mass, total uranium mass and identification for all declared UF 6 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 US Support Program team consisted of Pacific Northwest National Laboratory (PNNL, lead), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL) and Savanah River National Laboratory (SRNL). At the core of the viability study is a long-term field trial of a prototype UCVS system at a Westinghouse fuel fabrication facility. A key outcome of the study is 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 report provides context for the UCVS concept and the field trial: potential UCVS implementation concepts at an enrichment facility; an overview of UCVS prototype design; field trial objectives and activities. Field trial 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 the contents of a given cylinder are consistent with previous scans. A modeling study, combined with field-measured instrument uncertainties, provides an assessment of the partial-defect sensitivity of HEVA and PNEM for both one-time assay and (repeated) NDA Fingerprint verification scenarios. The findings presented in this report represent a significant step forward in the community’s understanding of the strengths and limitations of the PNEM and HEVA NDA methods, and the viability of the UCVS concept in front-end fuel cycle facilities. This experience will inform Phase II of the UCVS viability study, should the IAEA pursue it.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of a Novel Accelerator for Neutron Transport Solution Using the Galerkin Spectral Element Methods (Final Report)

This report summarizes the results of a three-year research project sponsored by the U.S. Department of Energy (DOE) Nuclear Energy University Program (NEUP) to develop and implement advanced acceleration schemes for the DOE NEAMS neutronics code PROTEUSSN. The project team included the University of Massachusetts Lowell, The Ohio State University, University of Michigan, and Argonne National Laboratory. The PROTEUS code is a high-fidelity capable deterministic neutron transport code based on unstructured finite element meshes, which solves the steady-state and transient neutron transport problem using the 2nd-order discrete ordinate method (SN2ND), the method of characteristics (MOC), and the advanced nodal transport method (NODAL). The DSA scheme has been implemented in PROTEUS to speed up SN transport calculations. The existing DSA scheme employs a consistent finite element formulation on the same fine mesh structure as the SN solution While the use of consistent discretizations makes the DSA method effective, it does not necessarily make it efficient because the numerical solution of a discretized elliptic diffusion problem itself can be costly, particularly when the problem size becomes large. We developed a new discontinuous Galerkin (DG) discretization of the diffusion equation, called DG-DSA, which can effectively and efficiently accelerate the SN transport iterations. As compared with the previous work, the novelty of our method is that the diffusion equation is solved on a coarse-mesh grid using the DG methods, and the DG diffusion discretization incorporates local hp adaptation, i.e., local adaptation of mesh size and/or polynomial degree, based on local total cross section (or optical thickness). Therefore, the resulting number of degrees of freedom (DOF) of the DG discretization is much less than the conventional consistent DSA discretizations, and thus DG-DSA can achieve significant improvement in computational efficiency. We implemented this scheme in PROTEUS-SN. In addition, we developed a new stabilization scheme named linear prolongation CMFD (lpCMFD). A novel feature of this scheme is that the conventional flat flux ratio–based scaling approach is replaced with a linear interpolation of the scalar flux differences at the coarse-mesh cell edges between the neutron transport and CMFD calculations. Fourier convergence analysis and numerical results show that lpCMFD is unconditionally stable and more effective than other CMFD based on acceleration schemes such as pCMFD, odCMFD. We also developed a new nonlinear diffusion acceleration scheme for solving neutron transport equations. This scheme, called LR-NDA, employs a local refinement approach on the framework of CMFD by solving a local boundary value problem of the scalar flux on the coarse-mesh structure to replace the piecewise constant scalar flux obtained by CMFD. The refined flux is then used to update the scalar flux in the neutron transport source iteration. It has been demonstrated that that LR-NDA is much more effective and stable than CMFD for a wide range of optical thicknesses. LR-NDA is a local adaptive method, which means LR-NDA does not necessarily require local refinement for all the coarse-mesh cells on the problem domain, i.e., it can be used only for relatively optically thick regions where the standard CMFD scheme would encounter the convergence problem.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1995

A computational environment for exhaust nozzle design

The Nozzle Design Associate (NDA) is a computational environment for the design of jet engine exhaust nozzles for supersonic aircraft. NDA may be used either to design new aircraft or to design new nozzles that adapt existing aircraft so they may be reutilized for new missions. NDA was developed in a collaboration between computer scientists at Rutgers University and exhaust nozzle designers at General Electric (GE) Aircraft Engines and General Electric Corporate Research and Development. The NDA project has two principal goals: to provide a useful engineering tool for exhaust nozzle design, and to explore fundamental research issues that arise in the application of automated design optimization methods to realistic engineering problems.

Gelsey, Andrew↗
DOE OSTI2019

Expanded Use of ISOCS Characterization Techniques for WIPP Certification at the Idaho Clean-up Project - 19623

The Advanced Mixed Waste Treatment Project (AMWTP) has been retrieving, treating, characterizing, and shipping legacy Transuranic (TRU) waste for disposal since 2003 as part of the overall Idaho Cleanup Project (ICP) in support of the US Department of Energy. As the ICP has progressed, over 700 containers have been identified that cannot be successfully assayed using the current suite of non-destructive assay (NDA) instruments. To meet this challenge, Fluor Idaho, LLC, and Mirion Technologies (formerly Canberra Industries) have teamed to deploy the In-Situ Object Counting System (ISOCS) to supplement the current NDA capabilities. ISOCS has long been used in various radioactive waste management applications, but never has the ISOCS ability to perform intrinsic efficiency modeling of discrete waste containers been employed to characterize and certify TRU waste for disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. Since the primary disposal facility for this waste is the WIPP, the characterization processes for determining the quantity of radioactive material entrained in the waste are subject to WIPP Waste Acceptance Criteria (WAC) quality requirements and associated performance criteria. Until now, the Carlsbad Field Office (CBFO) WIPP-certified NDA equipment has utilized pre-defined efficiency calibrations for fixed geometries. However, due to the radioactive material content, activity levels, waste stream matrix, and/or the geometries of various size drums, boxes and other-odd-shaped containers or container integrity issues, it may not be suitable to characterize these containers with any of the existing WIPP-certified NDA equipment located at the AMWTP. Therefore, advanced technologies are needed to complete characterization of these problematic waste containers. The July 5, 2016 release of the WIPP WAC, DOE/WIPP-02-3122, Revision 8.0, Appendix A, 'Radioassay Requirements for Transuranic Waste,' was expanded to include several newer concepts such as American Society for Testing and Materials (ASTM) approved methods for efficiency-based calibrations and the use of mathematical computer modeling techniques for calibration. This change prompted AMWTP to pursue obtaining WIPP certification authority for expanded ISOCS techniques to deal with the remaining challenging waste containers and the development and implementation of Mirion's Figures of Merit (FOM) evaluations. These vital FOM evaluations enhance the ISOCS's modeling capabilities by optimizing the dynamic efficiency calibration for each object and reducing the assay's Total Measurement Uncertainty (TMU) budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Passive gamma analysis of the boiling-water-reactor assemblies

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) 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. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137 Cs, 154 Eu, 134Cs, and to a lesser extent, 106 Ru and 144 Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Improved uncertainty quantification in nondestructive assay for nonproliferation

This paper illustrates methods to improve uncertainty quantification (UQ) for non-destructive assay (NDA) measurements used in nuclear nonproliferation. First, it is shown that current bottom-up UQ applied to calibration data is not always adequate, for three main reasons: (1) Because there are errors in both the predictors and the response, calibration involves a ratio of random quantities, and calibration data sets in NDA usually consist of only a modest number of samples (3–10); therefore, asymptotic approximations involving quantities needed for UQ such as means and variances are often not sufficiently accurate; (2) Common practice overlooks that calibration implies a partitioning of total error into random and systematic error, and (3) In many NDA applications, test items exhibit non-negligible departures in physical properties from calibration items, so model-based adjustments are used, but item-specific bias remains in some data. Therefore, improved bottom-up UQ using calibration data should predict the typical magnitude of item-specific bias, and the suggestion is to do so by including sources of item-specific bias in synthetic calibration data that is generated using a combination of modeling and real calibration data. Second, for measurements of the same nuclear material item by both the facility operator and international inspectors, current empirical (top-down) UQ is described for estimating operator and inspector systematic and random error variance components. A Bayesian alternative is introduced that easily accommodates constraints on variance components, and is more robust than current top-down methods to the underlying measurement error distributions.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Using the JCC-71 neutron coincidence collar as a benchmark for detector characterization with PTR-32 list mode data acquisition

Augmenting or interchanging current international safeguards shift registerbased data acquisition systems with list mode data acquisition systems gives nondestructive assay (NDA) systems greater versatility. Neutron list mode data analysis offers comparable analytical results to the more widely used shift register analysis in nuclear material quantification applications and offers several diagnostic tools that are specifically beneficial to in-laboratory characterization and calibration measurements. These benefits include identification of non-ideal behavior, optimization of operational parameters from a single measurement, and an improved understanding of the physics-based behavior of NDA systems for a more precise system representation and more confident assay results. In this work the advantages of using list mode data acquisition for detector characterization are demonstrated experimentally. Two commercial-off-the-shelf International Atomic Energy Agency-supported technologies are used for a typical detector characterization procedure and their performance is compared. Specifically, a 3 He-based Canberra Industries JCC-71 Neutron Coincidence Collar is characterized using the Hungarian Institute of Isotopes’ Pulse Train Recorder-32 (PTR-32) list mode data acquisition system, and the results are compared to those obtained using the standard Canberra Industries JSR-15 model shift register. The quantitative results from the two systems are in agreement, which demonstrates that the PTR-32 is a technically viable alternative to conventional shift register electronics for this task. A suitable procedure for full instrument characterization is described, and the added benefits of list mode for characterization and data collection are discussed. Furthermore, this is an important step toward establishing a procedure for the routine use of list mode data acquisition and analysis for neutron NDA system characterization in safeguards field applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Distinguishing fissile uranium isotopes using an active well neutron coincidence counter

Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

OR14-V-Uncertainty-PD2La Uncertainty Quantification for Nuclear Safeguards and Nondestructive Assay Final Report

The various methods of nondestructive assay (NDA) of special nuclear material (SNM) have applications in nuclear nonproliferation, including detection and identification of illicit SNM at border crossings and quantifying SNM at nuclear facilities for safeguards. No assay method is complete without “error bars,” which provide one way of expressing confidence in the assay result. Consequently, NDA specialists typically provide error bars and also partition total uncertainty into “random” and “systematic” components so that, for example, an error bar can be developed for the total mass estimate in multiple items. Uncertainty Quantification (UQ) for NDA has always been important, but it is recognized that greater rigor is needed and achievable using modern statistical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

ARIES Blend Lot Book Data Analysis

The Advanced Recovery Integrated Extraction System (ARIES) Program being executed at Los Alamos TA-55 produces unpolished plutonium oxide powder from pit disassembly in batches of material referred to as a blend lot. Each blend lot is sampled and analyzed by Los Alamos group C-AAC (Chemistry Division, Actinide Analytical Chemistry Group) for radioactive and nonradioactive constituents contained in the material. A blend lot, which nominally consists of 12 kg(Pu), is then packaged into three DOE-STD-3013 containers that are analyzed by nondestructive analysis (NDA) to determine the fraction and total amount of plutonium mass relative to impure plutonium oxide. The analytical and NDA data are presented in a Blend Lot Book (BLB). The data are used to confirm that a blend lot grouping of three cans meets the requirements for off-site shipment specified in the formal interface control document (ICD). The analytical and NDA data from 49 blend lots were reviewed to provide summary-level information on the constituents present in the blend lots. The results provide an overall perspective in historical amounts for constituent averages over all blend lots and similar information in the variation between constituent levels. Note that to properly compare the can contents, it would be most appropriate to decay all of cans’ constituents to a particular date. Previously, this analysis only examined statistical data as determined at the time of blend lot chemical analysis. The chemical analysis dates span the period from June 2011 to May 2015 (4 years), which is 28% of one Pu-241 half-life. Therefore, the approximate nature of the aggregate information must be properly understood in this light. However, to address this issue, a new table is included in this revision that decays the 48 blend lots to a single date.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Magnetic Microcalorimeter Gamma-Detectors for High-Accuracy Nuclear Decay Data

Magnetic microcalorimeters (MMCs) are low-energy gamma-ray detectors with ultra-high energy resolution that are operated at temperatures below 50 mK. They use metallic Au absorbers and paramagnetic Ag:Er sensors to measure gamma-ray energies from an increase in detector temperature. Ultra-low temperatures are made accessible to non-expert users with “dry” dilution refrigerators that replace the traditional liquid N 2 and He pre-cooling with a mechanically cooled pulse-tube refrigerator. Our MMC gamma detectors have achieved an energy resolution below 40 eV FWHM at 60 keV, and they have a uniform and reproducible response function that allows adding spectra from different detector pixels without loss in energy resolution. These properties make MMCs well-suited to re-measure nuclear decay data with increased precision and smaller systematic errors to increase the accuracy of non-destructive assays (NDA). We present recent MMC gamma-ray spectra of Th-234 to show how MMCs can resolve lines that are relevant for NDA analysis and that overlap in Ge spectra. We also discuss the accuracy that MMCs can achieve for nuclear decay data with an emphasis on gamma-ray energies and intensity ratios that current NDA codes are based on.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

NA-22 Write for Consortium Efforts

We are asked to look into the improvements for the measurement of Pu of a variety of material types (MT): MT41 (Pu-242, 20 – 60% enriched), MT42 (Pu-242 > 60% enrichment), and MT-82 (Np-237). Many of these are legacy items that were packaged years ago (Fig 1&2), and when the items are re-packaged for long-term storage, they must be re-measured using either NDA (nondestructive analysis methods or radiochemical methods. It is of interest to improve the NDA and radiochemistry to reduce the un-certainty in the measurements, as in the case of larger items (250 g to 2 kg items) the uncertainty in measurements makes it difficult to certify the mass of the material down to the 0.5 g accountability requirement per DOE orders by only doing NDA measurement. It would be of interest to look into methods to improve the detection of these materials, which would not only benefit the weapon manufacturing community, but also has the ability to significantly improve the non-proliferation community.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗