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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3 He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work covers the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Attributes from NMIS Time Coincidence, Fast-Neutron Imaging, Fission Mapping, And Gamma-Ray Spectrometry Data

This work tests a systematic procedure for analyzing data acquired by the Nuclear Materials Identification System (NMIS) at Oak Ridge National Laboratory with fast-neutron imaging and high-purity germanium (HPGe) gamma spectrometry capabilities. NMIS has been under development by the US Department of Energy Office of Nuclear Verification since the mid-1990s, and prior to that by the National Nuclear Security Administration Y-12 National Security Complex, with NMIS having been used at Y-12 for template matching to confirm inventory and receipts. In this present work, a complete set of NMIS time coincidence, fast-neutron imaging, fission mapping, and HPGe gamma-ray spectrometry data was obtained from Monte Carlo simulations for a configuration of fissile and nonfissile materials. The data were then presented for analysis to someone who had no prior knowledge of the unknown object to accurately determine the description of the object by applying the previously mentioned procedure to the simulated data. The best approximation indicated that the unknown object was composed of concentric cylinders: a void inside highly enriched uranium (HEU) (84.7 {+-} 1.9 wt % 235 U), surrounded by depleted uranium, surrounded by polyethylene. The final estimation of the unknown object had the correct materials and geometry, with error in the radius estimates of material regions varying from 1.58% at best and 4.25% at worst; error in the height estimates varied from 2% to 12%. The error in the HEU enrichment estimate was 5.9 wt % (within 2.5Σ of the true value). The accuracies of the determinations could be adequate for arms control applications. Future work will apply this iterative reconstructive procedure to other unknown objects to further test and refine it.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Recent Progress on the PNL Accelerator-Based Intense Fusion Neutron Source

Phoenix Nuclear Labs (PNL) has developed a full line of accelerator-based neutron generators. This paper describes the highest yield DD fusion-based system. This device utilizes a microwave ion source (MWS), 300 kV DC accelerator, magnetic solenoid focus element, differential pumping system, and gaseous deuterium target to achieve DD neutron yields of 3x10{sup 11} n/s. Lower-yield variations of the device have been built using variable voltages and a solid titanium target, and design for a DT version of the gas target system is underway that will increase neutron yield to 5x10{sup 13} n/s. PNL has delivered a number of systems to government and commercial customers and has identified a number of longer-term commercial applications for this high-yield neutron generator. These include medical isotope production, neutron radiography, active interrogation for explosives and SNM detection, and Cf-252 replacement. Most applications require the development of specialized moderator assemblies and fixtures to meet customer requirements. A photograph of the neutron generator prototype is provided in Figure 1. This state-of-the-art neutron generator technology is essential for many applications of critical global importance including neutron radiography, medical isotope production, detection of explosives and nuclear material, materials characterization, and others. The presentation will discuss these applications and recent progress made on the base accelerator technology. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
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 OSTI2012

Detecting fission from special nuclear material sources

A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a graphing component that displays the plot of the neutron distribution from the unknown source over a Poisson distribution and a plot of neutrons due to background or environmental sources. The system further includes a known neutron source placed in proximity to the unknown source to actively interrogate the unknown source in order to accentuate differences in neutron emission from the unknown source from Poisson distributions and/or environmental sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

An Evaluation of a Neutron Time Correlated Interrogation Method for Measurement of Fissile Content in Research Reactor Spent Fuel Assemblies

There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source ( 252 Cf) for spent fuel interrogations was evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Idaho Explosive Detection System

Learn how INL researchers are making the world safer by developing an explosives detection system that can inspect cargo. For more information about INL security research, visit http://www.facebook.com/idahonationallaboratory

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Development and evaluation of a list mode neutron coincidence collar for spatial response measurements of fresh fuel assemblies

A traditional safeguards neutron coincidence counting system, the Neutron Coincidence Collar, has been modified to incorporate individual preamplifiers on each of its 18 3He tubes in active interrogation mode. When used with list mode data acquisition (LMDA) and analysis, a signal from each 3He tube can be recorded and analyzed to allow a spatial response measurement to be performed on an item based on count rate and 3He tube location within the system. The ultimate goal of this project is to demonstrate the capabilities of a list mode response matrix for the nondestructive assay of fresh nuclear fuel assemblies. To enable partial defect detection of fuel pin locations and absences within an assembly, the project aims to extend well-established correlated neutron analysis techniques on a preexisting Neutron Coincidence Collar by extracting a greater number of useful signatures from the system than are currently generated. LMDA, combined with the addition of multiple preamplifiers, facilitates this capability by increasing the number of simultaneous signals that can be measured; this allows an in-depth analysis of neutron coincidence events to determine a fissioning item’s location based on the measured doubles count rate in various channel logic coincidence combinations. All of this can be done from a single measurement pulse train in offline analysis, which is the major benefit provided by LMDA. Through various stages of development and testing, a Mirion Technologies model JCC-71 Neutron Coincidence Collar has been successfully retrofit with modern electronics designed in-house at Oak Ridge National Laboratory, matching preexisting JAB-01 electronics performance, while maintaining the original system footprint. This paper presents these various stages of development and experimental evaluation of the proof of concept system.

Moore, Angela S.↗
DOE OSTI2015

Active Interrogation for Spent Fuel

The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Stilbene Scintillator Detector Array and Data-Acquisition System for Active Interrogation

A large area Stilbene crystal detector array has been assembled and calibrated for use in active interrogation measurements. The array consists of 10 high-efficiency 4”x2” stilbene cells. The cells were grown by Inrad Optics using LLNL developed technology. The array can be used to detect multiple signatures of SNM items, both actively and passively. Prompt and delayed gamma rays as well as prompt and delay neutrons can be detected and counted separately. The array is coupled to the Mesytec MDPP-16 VME-based digitizer module for data acquisition. This module performs on-board pulse-shape discrimination of signals from the Stilbene crystal detectors. This data-acquisition module allows separation of neutron and gamma-ray events with a low demand on data transfer rates. In addition, the high-resolution timing (<100 ps) is suitable for correlated gamma-ray and neutron multiplicity counting. This report describes the construction and calibration of the system. We also discuss applications of the system as well as how the system is currently being utilized by end users.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Multi-particle inspection using associated particle sources

Disclosed herein are representative embodiments of methods, apparatus, and systems for performing combined neutron and gamma ray radiography. For example, one exemplary system comprises: a neutron source; a set of alpha particle detectors configured to detect alpha particles associated with neutrons generated by the neutron source; neutron detectors positioned to detect at least some of the neutrons generated by the neutron source; a gamma ray source; a set of verification gamma ray detectors configured to detect verification gamma rays associated with gamma rays generated by the gamma ray source; a set of gamma ray detectors configured to detect gamma rays generated by the gamma ray source; and an interrogation region located between the neutron source, the gamma ray source, the neutron detectors, and the gamma ray detectors.

42 ENGINEERING↗
DOE OSTI2010

Time-Spectral Analysis Methods for Spent Fuel Assay Using Lead Slowing-Down Spectroscopy of Spent Fuel

Nondestructive techniques for measuring the mass of fissile isotopes in spent nuclear fuel is a considerable challenge in the safeguarding of nuclear fuel cycles. A nondestructive assay technology that could provide direct measurement of fissile mass, particularly for the plutonium (Pu) isotopes, and improve upon the uncertainty of today’s confirmatory methods is needed. Lead slowing-down spectroscopy (LSDS) has been studied for the spent fuel application previously, but the nonlinear effects of assembly self shielding (of the interrogating neutron population) have led to discouraging assay accuracy for realistic pressurized water reactor fuels. In this paper, we describe the development of time-spectral analysis algorithms for LSDS intended to overcome these self-shielding effects. The algorithm incorporates the tabulated energy-dependent cross sections from key fissile and absorbing isotopes, but leaves their mass as free variables. Multi-parameter regression analysis is then used to directly calculate not only the mass of fissile isotopes in the fuel assembly (e.g., Pu-239, U-235, and Pu-241), but also the mass of key absorbing isotopes such as Pu-240 and U-238. Modeling-based assay results using a first-order self-shielding relationship indicate that LSDS has the potential to directly measure fissile isotopes with less than 5% average relative error, over a wide fuel burnup range. Shortcomings in the first-order self-shielding model and methods to improve the formulation are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Combined Assay Systems for the Characterisation of Historical ILW at of the Dounreay Shaft and Silo-18140

In 1955 the Dounreay nuclear power development establishment was created on the northern coast of Scotland primarily to pursue the UK Government policy of developing Fast Breeder Reactor (FBR) technology. The site operated then by the United Kingdom Atomic Energy Authority (UKAEA) saw the development, test and operation of 3 nuclear reactors, two of them FBRs, alongside fabrication and reprocessing facilities for the FBR fuel elements. From the late 1950's until 1977 Intermediate Level radioactive Waste (ILW) generated on site was disposed of into a 65 m deep, 4.6 m diameter shaft. In January 1977, an explosion thought to be caused by hydrogen generated from sodium and potassium wastes reacting with water occurred in the upper part of the shaft, thus stopping any further waste disposal. A second ILW storage facility, the ILW-Silo, started operation in 1971 to deal with the increasing volume of intermediate level waste generated on site in the mid-1960's and continued receiving waste until 1998. The decommissioning of the Shaft and Silo facilities is a challenging engineering task due to the complexity and potential hazards linked with the retrieval and handling of historical waste. Following recovery it is intended to shred the majority of the waste and load it into 200-l assay drums. Sludges would be separated and solidified within a special annulus drum to make them suitable for assay. To support the decommissioning strategies put forward by Dounreay Site Restoration Limited (DSRL), Mirion Technologies Canberra UK Ltd have designed and completed the manufacture of a pair of combined non-destructive waste assay (NDA) systems. The two identical automated ILW counters are intended to provide a means to characterise the waste materials by performing active and passive neutron measurement as well as segmented high resolution gamma spectrometry (HRGS) of 200-l sacrificial waste drums. The active neutron technique is based on the Differential Die Away (DDA) technique using a pulsed Deuterium-Tritium neutron generator to interrogate fissile material present in the waste. The active neutron measurements support, as well as the passive neutron coincidence measurements, a matrix correction capability by use of an Add-a-Source mechanism. The publication will describe in broad lines the approach proposed by DSRL in the processing of the waste retrieved from both the Shaft and Silo facilities. It will also present a description of the overall design of the NDA counters, illustrating how the measurements to be performed will allow DSRL to safely characterise the retrieved waste materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Optimization of combined delayed neutron and differential die-away prompt neutron signal detection for characterization of spent nuclear fuel assemblies

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy (DOE) has funded multiple laboratories and universities to develop a means to accurately quantify the Plutonium (Pu) mass in spent nuclear fuel assemblies and ways to also detect potential diversion of fuel pins. Delayed Neutron (DN) counting provides a signature somewhat more sensitive to {sup 235}U than Pu while Differential Die-Away (DDA) is complementary in that it has greater sensitivity to Pu. The two methods can, with care, be combined into a single instrument which also provides passive neutron information. Individually the techniques cannot robustly quantify the Pu content but coupled together the information content in the signatures enables Pu quantification separate to the total fissile content. The challenge of merging DN and DDA, prompt neutron (PN) signal, capabilities in the same design is the focus of this paper. Other possibilities also suggest themselves, such as a direct measurement of the reactivity (multiplication) by either the boost in signal obtained during the active interrogation itself or by the extension of the die-away profile. In an early study, conceptual designs have been modeled using a neutron detector comprising fission chambers or 3He proportional counters and a {approx}14 MeV neutron Deuterium-Tritium (DT) generator as the interrogation source. Modeling was performed using the radiation transport code Monte Carlo N-Particles eXtended (MCNPX). Building on this foundation, the present paper quantifies the capability of a new design using an array of {sup 3}He detectors together with fission chambers to optimize both DN and PN detections and active characterization, respectively. This new design was created in order to minimize fission in {sup 238}U (a nuisance DN emitter), to use a realistic neutron generator, to reduce the cost and to achieve near spatial interrogation and detection of the DN and PN, important for detection of diversion, all within the constraints of a single practical instrument. Both DN and PN detections are active techniques using the signal from the most prominent fissile isotopes of spent nuclear fuel that respond the best to a slow neutron interrogation, {sup 235}U, {sup 239}U and {sup 241}PU. The performance is characterized against a library of 64 assemblies and 40 diversion scenarios at different burnup (BU), cooling-time (CT) and initial enrichment (IE) in fresh water.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗