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Alexis C. Kaplan

Other affiliations: Los Alamos National Laboratory
Bio: Alexis C. Kaplan is an academic researcher from University of Michigan. The author has contributed to research in topics: Spent nuclear fuel & Burnup. The author has an hindex of 5, co-authored 9 publications receiving 79 citations. Previous affiliations of Alexis C. Kaplan include Los Alamos National Laboratory.

Papers
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Journal ArticleDOI
TL;DR: In this paper, the EJ-309 scintillator was used to measure neutrons in the presence of high gamma-ray fluence, which is a challenge with multi-particle detectors.
Abstract: Measuring neutrons in the presence of high gamma-ray fluence is a challenge with multi-particle detectors. Organic liquid scintillators such as the EJ-309 are capable of accurate pulse-shape discrimination (PSD) but the chance for particle misclassification is not negligible for some applications. By varying the distance from an EJ-309 scintillator to a strong-gamma-ray source and keeping a weak-neutron source at a fixed position, various gamma-to-neutron ratios can be measured and PSD performance can be quantified. Comparing neutron pulse-height distributions allows for pulse-height specific PSD evaluation, and quantification and visualization of deviation from 252Cf alone. Even with the addition of the misclassified gamma-rays, the PSD is effective in separating particles so that neutron count rate can be predicted with less than 10% error up to a gamma-to-neutron ratio of almost 650. For applications which can afford a reduction in neutron detection efficiency, PSD can be sufficiently effective in discriminating particles to measure a weak neutron source in a high gamma-ray background.

34 citations

Journal ArticleDOI
TL;DR: In this paper, a differential die-away self-interrogation (DDSI) was used to determine the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation instrument.
Abstract: We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four 3He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

15 citations

ReportDOI
01 Oct 2012
TL;DR: In this paper, the authors examined design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes.
Abstract: This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement more » durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" O ? 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" O ? 3" detectors. « less

10 citations

01 Jan 2014
TL;DR: In this article, the capability of non-destructive assay (NDA) techniques to meet the combined needs of the safeguards community and the Swedish encapsulation and repository facilities operator SKB was evaluated.
Abstract: The Swedish Nuclear Fuel and Waste Management Company (SKB), European Atomic Energy Community (Euratom), two universities and several U.S. Department of Energy Laboratories have joined in a collaborative research effort to determine the capability of non-destructive assay (NDA) techniques to meet the combined needs of the safeguards community and the Swedish encapsulation and repository facilities operator SKB. These needs include partial defect detection, heat quantification, assembly identification (initial enrichment, burnup and cooling time), and Pu mass and reactivity determination. The experimental component of this research effort involves the measurement of 50 assemblies at the Central Storage of Spent Nuclear Fuel (Clab) facility in Sweden, 25 of which were irradiated in Pressurized Water Reactors and 25 in Boiling Water Reactors. The experimental signatures being measured for all assemblies include spectral resolved gammas (HPGe and LaBr3), time correlated neutrons (Differential Die-away Self Interrogation), time-varying and continuous active neutron interrogation (Differential Die-away and an approximation of Californium Interrogation Prompt Neutron), total neutron and total gamma fluxes (Fork Detector), total heat (assembly length calorimeter) and possibly the Cerenkov light emission (Digital Cerenkov Viewing Device). This paper fits into the IAEA’s Department of Safeguards Long-Term R&D Plan in the context of developing “more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect test on spent fuel assembly prior to transfer to difficult to access storage,” as well as potentially supporting pyrochemical processing. The work describes the specific measured signatures, the uniqueness of the information contained in these signatures and why a data mining approach is being used to combine the various signatures to optimally satisfy the various needs of the collaboration. This paper will address efficient and effective verification strategies particularly in the context of encapsulation and repository facilities.

10 citations

Journal ArticleDOI
TL;DR: In this article, the authors used differential die-away self-interrogation (DDSI) to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA).
Abstract: As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

7 citations


Cited by
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Journal ArticleDOI
TL;DR: In this article, a solution-based technique for growing large-volume stilbene scintillators was developed in 2013; crystals up to diameters of 10 cm, or larger, have been grown while preserving excellent pulse shape discrimination (PSD) properties.
Abstract: A solution-based technique for growing large-volume stilbene scintillators was developed in 2013; crystals up to diameters of 10 cm, or larger, have been grown while preserving excellent pulse shape discrimination (PSD) properties. The goal of this study is to evaluate the PSD capabilities of 5.08 by 5.08-cm stilbene crystals grown by Lawrence Livermore National Laboratory and Inrad Optics when exposed to a 1000 to 1 gamma ray-neutron ratio and operating at a 100-kHz count rate. Results were compared to an equivalent EJ-309 liquid scintillation detector. 252 Cf neutron pulses were recorded in two experiments where 60 Co and 137 Cs sources created the high-gamma field. The high count rate created numerous double pulses that were cleaned using fractional and template approaches designed to remove double pulses while preserving neutron counts. PSD was performed at a threshold of 42 keVee (440-keV proton) for stilbene and 60 keVee (610-keV proton) for EJ-309 liquid. The lower threshold in stilbene resulted in a neutron intrinsic efficiency of approximately 14.5%, 10% higher than EJ-309 liquid, for bare 252 Cf and 13% for 252 Cf in the high-gamma field. Despite the lower threshold, the gamma misclassification rate in stilbene was approximately 3×10 −6 , nearly a factor-of-five lower than what we found with the EJ-309 liquid.

46 citations

Journal ArticleDOI
TL;DR: The Next Generation Safeguards Initiative (NGSI) -Spent Fuel (SF) project as discussed by the authors has developed a set of measurement campaigns at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB).
Abstract: The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF 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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

34 citations

Journal ArticleDOI
TL;DR: In this article, a fast-neutron multiplicity counter developed at the University of Michigan was used to detect neutrons in Ispra, Italy, where the measurements allowed the assessment of the system's photon discrimination abilities, efficiency when measuring neutron multiplicity, ability to characterize 240Pueff mass, and performance relative to a currently deployed neutron coincidence counter.
Abstract: Measurements were performed at the Joint Research Centre in Ispra, Italy to field test a fast-neutron multiplicity counter developed at the University of Michigan. The measurements allowed the assessment of the system׳s photon discrimination abilities, efficiency when measuring neutron multiplicity, ability to characterize 240Pueff mass, and performance relative to a currently deployed neutron coincidence counter. This work is motivated by the need to replace and improve upon 3He neutron detection systems for nuclear safeguards applications.

28 citations

Journal ArticleDOI
TL;DR: In this article, the expected performance of a conceptual fast-neutron multiplicity system (FNMS) for assaying plutonium was evaluated using Monte Carlo simulations, and the results from these simulations were then used to estimate the Pu assay capabilities of the FNMS in terms of counting time, assay mass, and assay mass variance.
Abstract: Statistical analyses have been performed to develop bounding estimates of the expected performance of a conceptual fast-neutron multiplicity system (FNMS) for assaying plutonium. The conceptual FNMS design includes 32 cubic liquid scintillator detectors, measuring 7.62 cm per side, configured into 4 stacked rings of 8 detectors each. Expected response characteristics for the individual FNMS detectors, as well as the response characteristics of the entire FNMS, were determined using Monte Carlo simulations based on prior validation experiments. The results from these simulations were then used to estimate the Pu assay capabilities of the FNMS in terms of counting time, assay mass, and assay mass variance, using assay mass variance as a figure of merit. The analysis results are compared against a commonly used thermal-neutron coincidence counter. The advantages of using a fast-neutron counting system versus a thermal-neutron counting system are significant. Most notably, the time required to perform an assay to an equivalent assay mass variance is greatly reduced with a fast-neutron system, by more than an order of magnitude compared with that of the thermal-neutron system, due to the reduced probability of random summing with the fast system. The improved FNMS performance is especially relevant for assays involving Pu masses of 10 g or more.

27 citations