Nondestructive Assay Instruments Research Articles

Identification of diversions in spent PWR fuel assemblies by PDET signatures using Artificial Neural Networks (ANNs)

Spent nuclear fuel represents the majority of materials placed under nuclear safeguards today and it requires to be inspected and verified regularly to promptly detect any illegal diversion. Research is ongoing both on the development of non-destructive assay instruments and methods for data analysis in order to enhance the verification accuracy and reduce the inspection time. In this paper, two models based on Artificial Neural Networks (ANNs) are studied to process measurements from the Partial Defect Tester (PDET) in spent fuel assemblies of Pressurized Water Reactors (PWRs), and thus to identify at different levels of detail whether nuclear fuel has been replaced with dummy pins or not. The first model provides an estimation of the percentage of replaced fuel pins within the inspected fuel assembly, while the second model determines the exact configuration of the replaced fuel pins. The two models are trained and tested using a dataset of Monte-Carlo simulated PDET responses for intact spent PWR fuel assemblies and a variety of hypothetical diversion scenarios. The first model classifies fuel assemblies according to the percentage of diverted fuel with a high accuracy (96.5%). The second model reconstructs the correct configuration for 57.5% of the fuel assemblies available in the dataset and still retrieves meaningful information of the diversion pattern in many of the misclassified cases.

Modernizing the active neutron collar poison rod correction using simulations to enhance PWR fresh fuel verification

The declared inventory of 235U in fresh, low enriched uranium, nuclear fuel assemblies is routinely verified by nuclear safeguards inspectorates using the UNCL (Uranium Neutron Collar – Light Water Reactor Fuel) nondestructive assay instrument. The trend in modern fuels is towards higher initial enrichment, which in turn requires a larger number of burnable poison pins with higher Gadolinia concentration to hold down the initial reactivity. UNCL assay error for modern fuel assemblies typically exceeds 10%. The traditional algorithm used to correct the response for burnable poison content needs revision to achieve accuracy comparable to the σR of 4.5% expected for poison-free assemblies. In this paper we review the problem and use a large set of 287 Monte Carlo simulations based on an experimentally benchmarked model corresponding to the 16×16 PWR array used in Brazilian Angra type II and III fuel. Simulated relative responses are used to evaluate the functional form of the poison correction and determine parameters optimal for 16×16 PWR assemblies. This update is conducted using between 4–24 Gadolinia pins ranging 2–11 wt% Gd2O3 with assemblies having mean enrichments ranging 2.5%–5%. Experimental validation of the updated coefficients using seven measurements shows bias is reduced by an order of magnitude and σR reduced from 10% to less than 2%. Two updated sets of coefficients are provided, the most accurate of which is directly useable in the existing analysis code (INCC) in use by inspectorate at fuel fabrication facilities.

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

Nondestructive measurements of residual 235U mass of Israeli Research Reactor-1 fuel using the Advanced Experimental Fuel Counter

In 2018, a measurement campaign took place with participants from Los Alamos National Laboratory (LANL), the Nuclear Research Centre-Negev (NRCN) and Soreq Nuclear Research Center (SNRC) at the Israeli Research Reactor-1 (IRR-1) in which 14 of the reactor’s used fuel assemblies (FAs) with varied amount of depletion were measured with the nondestructive assay instrument Advanced Experimental Fuel Counter (AEFC). Designed for safeguards purposes, the AEFC measures both neutrons emitted from the FA (passive neutrons) and fission neutrons induced by an external neutron source (in this experiment, 252Cf). Signals recorded with the AEFC include total neutron count rates (Singles), time-correlated neutron count rates (Doubles), and total gamma-ray count rates. The 235U content of the FAs was previously assessed by two independent methods: (1) measurement of the transparency of the FA to low-energy gamma rays from an activated rhenium source (rhenium gamma transmission, or the RGT method) and (2) calculation of the ∼30-year burnup history of the core using detailed three-dimensional Monte-Carlo core depletion calculations. The results from the FAs that had been measured via the RGT method were used to construct the calibration curves, which translate the AEFC count rates to 235U mass. Then, the calibration was evaluated using AEFC measurements of six additional FAs that were not measured via the RGT method. From the results, it was determined the Doubles calibration curve was more reliable than that of the Singles and follows a simple second-order polynomial fit for the whole range of residual 235U mass content, albeit with larger statistical uncertainty.Detailed uncertainties quantification was conducted for both the AEFC Singles and Doubles. This includes the analysis of statistical uncertainties, calibration uncertainty, and random uncertainties due to the sensitivity of the AEFC to several sources of uncertainty, namely the FA position, FA orientation, interrogation source position, and ambient pool temperature. An overall total uncertainty of 6 g of 235U is estimated for the Singles and Doubles, which is mainly due to calibration uncertainty (for the Singles) and statistical uncertainty (for the Doubles), and which constitutes 3%–6% of the 235U total mass in the FAs, depending on their level of depletion.

Results of the Swedish spent fuel measurement field trials with the Differential Die-Away Self-Interrogation Instrument

Differential Die-Away Self-Interrogation (DDSI) is a method by which the characteristic die-away of neutrons from spontaneous and induced fissions is used to characterize a spent nuclear fuel assembly. A nondestructive assay (NDA) instrument was built at Los Alamos National Laboratory to implement and test the DDSI method. The DDSI instrument contains 3He detectors which measure thermal neutrons, and the time and location of detection of each neutron is recorded via list-mode data acquisition. The instrument was sent to the Clab interim storage facility in Sweden for measurement and characterization of 50 spent pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. The result was over 40 h of neutron list-mode data from a wide variety of fuel assemblies with high enough efficiency to perform neutron coincidence counting, i.e. detection of time-correlated neutrons from fission. Analysis algorithms for characterization of the fuel assemblies were tested on the Swedish spent fuel dataset. Using the measured data, multiplication, fissile mass, initial enrichment, burnup, and total plutonium mass were determined in the 50 assemblies with root mean square errors ranging from 1.5% for PWR multiplication to 11.4% for BWR fissile mass. The results in this work demonstrate that the DDSI concept is capable of characterizing spent power reactor fuel with levels of accuracy that are compatible with the requirements and objectives of various applications such as safeguards verification or facility material control and accounting.

Optimization of single-photon emission computed tomography system for fast verification of spent fuel assembly: a Monte Carlo study

The International Atomic Energy Agency (IAEA) assessed that single-photon emission computed tomography (SPECT) technique is one of the most attractive technique for safeguards of a spent fuel because of its intuitive evaluation capability by using a tomographic image. Even though there are over a hundred kinds of the verification technique, the IAEA suggested the need for “more sensitive and less intrusive alternatives to existing nondestructive assay instruments” for partial-defect detection. The aim of this study is to optimize a SPECT system using Monte Carlo method for faster verification of spent fuel assembly than the conventional system. An optimization study of detector geometry was performed using GATE simulation program. The detector head consists of one-dimensional (1D) multi-slit tungsten collimator, Bismuth Germanate (BGO) scintillator, and silicon photomultiplier (SiPM) arrays. The collimator slit width and length, and scintillator length were optimally determined by evaluating the quality of 1D projection image for Cs-137 line source. The type of photoelectric sensor was determined by assessing the light transfer efficiency of the scintillator using DETECT2000 program. For the performance evaluation of the optimized detector, a tomographic image of twelve fuel sources in the assembly was compared with that acquired using the conventional detector developed by IAEA . The optimized scintillator length, collimator slit width, length, and septal thickness were 4 cm, 0.2 cm, 5 cm, and 0.2 cm, respectively. The light transfer efficiency in the scintillator to the 3×70 mm3 and 3×3 mm2 sensors were 23.4± 0.6% and 5.3± 0.3%, respectively. In the results of the image quality assessment, the optimized detector head shows about 1.5 times improved sensitivity, while the image quality is slightly poorer than the conventional one due to the bigger slit width. Nevertheless, by considering a deep-learning-based image-reconstruction-algorithm, we chose the bigger slit width because higher image intensity gives an advantage in discrimination between the fuel sources and the background noise. We expect that the combination of the detector head with higher sensitivity and the deep-learning-based image-reconstruction-algorithm will contribute to the faster verification of spent fuel assemblies compared to the conventional system.

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 137Cs, 154Eu, and 134Cs 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. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

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: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. 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.

Design and performance of A 3He-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. 3He-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 3He gas was initiated worldwide. As part of this effort, Los Alamos National Laboratory (LANL) designed and built a 3He-free full scale thermal neutron coincidence counter based on boron-lined proportional technology. The boron-lined 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 3He-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.

Traceable Determination of the Absolute <newline/>Neutron Emission Yields of <formula formulatype="inline"><tex Notation="TeX">${{\rm UO}_2}{{\rm F}_2}$</tex></formula> <newline/>Working Reference Materials

The nuclear material contained in the process equipment of a uranium enrichment plant (referred to as holdup) is an important component of the overall nuclear material inventory for the plant. Accurate quantification and verification of holdup is needed to improve international safeguards and nuclear material accountancy. This is also needed for criticality safety and waste disposition. Passive neutron and gamma-ray nondestructive assay (NDA) methods are used to measure the holdup in process equipment. A key advantage of neutron measurements is that neutrons are highly penetrating and can be measured through thick walled equipment. The dominant source of neutrons in the UO2F2 holdup is from the 19F(α,n)22Na reaction resulting from 234U alpha decay when uranium is enriched. There is a considerable spread between different historic determinations of the 19F(α,n) yield from uranium which limits the accuracy of modeling and the calibration of NDA instruments. Furthermore, the compound form and presence of water also significantly affects the neutron emission rate from the holdup. This paper describes a series of experimental measurements performed at Los Alamos National Laboratory (LANL) to determine the absolute neutron emission yield from 10 different UO2F2 working reference materials (WRMs) fabricated at the Portsmouth Gaseous Diffusion Plant (PGDP). The Mini Epithermal Neutron Multiplicity Counter (Mini ENMC) and a NIST certified 252Cf neutron source were used for these measurements. The high efficiency and short die-away time of the Mini ENMC provides the high measurement precision needed to certify the neutron emission yield. The experiment was designed to achieve sub 1% accuracy in the net counting rate on each item and to provide assurance that important factors such as instrument stability, item placement and background were well understood. The traceable neutron yields measured from the WRMs were used to determine a more accurate neutron yield for the UO2F2 material. The results were compared to historical neutron emission rates. The F(α,n) data obtained from these measurements directly supports nuclear safeguards for NDA of uranium holdup and provides a more accurate calibration for new and existing NDA detector systems.

Development of an analytical theory to describe the PNAR and CIPN nondestructive assay techniques

This paper develops an analytical theory to describe how neutron albedo (reflection) increases the multiplication of neutrons by a used fuel assembly. With this theory, the two nondestructive assay (NDA) techniques of Passive Neutron Albedo Reactivity (PNAR) and Californium-252 Interrogation with Prompt Neutron Detection (CIPN) can be compared directly. Specifically, the theory derives expressions for the PNAR and CIPN metrics in terms of the physical properties of the used fuel assembly, such as the neutron multiplications and fate probabilities. The theory thus clarifies the interpretation of these two NDA techniques and suggests ways to improve both the design of the NDA instruments and the algorithms for analyzing the measurement results.

Calibration of Nondestructive Assay Instruments: An Application of Linear Regression and Propagation of Variance

Several nondestructive assay (NDA) methods to quantify special nuclear materials use calibration curves that are linear in the predictor, either directly or as an intermediate step. The linear response model is also often used to illustrate the fundamentals of calibration, and is the usual detector behavior assumed when evaluating detection limits. It is therefore important for the NDA community to have a common understanding of how to implement a linear calibration according to the common method of least squares and how to assess uncertainty in inferred nuclear quantities during the prediction stage following calibration. Therefore, this paper illustrates regression, residual diagnostics, effect of estimation errors in estimated variances used for weighted least squares, and variance propagation in a form suitable for implementation. Before the calibration can be used, a transformation of axes is required; this step, along with variance propagation is not currently explained in available NDA standard guidelines. The role of systematic and random uncertainty is illustrated and expands on that given previously for the chosen practical NDA example. A listing of open-source software is provided in the Appendix.

Measured F(α,n) Yield from 234U in Uranium Hexafluoride

As new uranium enrichment plants are proposed and come online worldwide, interest in using neutron methods for uranium hexafluoride (UF6) cylinder assay has been growing; however, large discrepancies exist in published F(α,n) yields from uranium isotopes. Uncertainties in these data are propagated through the analysis of every UF6 measurement and have implications for safeguards conclusions drawn from them. In this paper, a value for the specific F(α,n) yield in UF6 from 234U is calculated from measurements of 30B cylinders containing bulk UF6 at the Rokkasho Enrichment Plant in Japan. The measurements were taken using the Uranium Cylinder Assay System. The yield was derived by combining the cylinder measurements with detailed Monte Carlo modeling, known isotopic composition, and inversion analysis. We calculated the 234U neutron emission rate in UF6 to be (474 ± 21) n/s·g−1 with a 68% confidence level. The results obtained in this study will help enable an important class of nondestructive assay instruments to be applied with greater confidence and accuracy.

On Using Code Emulators and Monte Carlo Estimation to Predict Assembly Attributes of Spent Fuel Assemblies for Safeguards Applications

The quantification of the plutonium mass in spent nuclear fuel assemblies is an important measurement for nuclear safeguards practitioners. A program is well underway to develop nondestructive assay instruments that, when combined, will be able to quantify the plutonium content of a spent nuclear fuel assembly. Each instrument will quantify a specific attribute of the spent fuel assembly, e.g., the fissile content. In this paper, we present a Monte Carlo-based method of estimating the mean and distribution of some assembly attributes. An MCNPX model of each instrument has been created, and the response of the instrument was simulated for a range of spent fuel assemblies with discrete parameters (e.g., burnup, initial enrichment, and cooling time). The Monte Carlo-based method interpolates between the modeled results for an instrument to emulate a response for parameters not explicitly modeled. We demonstrate the usefulness of this technique in applying the technique to six different instruments under investigation. The results show that this Monte Carlo-based method can be used to estimate the assembly attributes of a spent fuel assembly based upon the measured response from the instrument.

Monte Carlo calculations of the neutron coincidence gate utilisation factor for passive neutron coincidence counting

The general purpose neutron–photon–electron Monte Carlo N-Particle code, MCNP TM , has been used to simulate the neutronic characteristics of the on-site laboratory passive neutron coincidence counter to be installed, under Euratom Safeguards Directorate supervision, at the Sellafield reprocessing plant in Cumbria, UK. This detector is part of a series of nondestructive assay instruments to be installed for the accurate determination of the plutonium content of nuclear materials. The present work focuses on one aspect of this task, namely, the accurate calculation of the coincidence gate utilisation factor. This parameter is an important term in the interpretative model used to analyse the passive neutron coincidence count data acquired using pulse train deconvolution electronics based on the shift register technique. It accounts for the limited proportion of neutrons detected within the time interval for which the electronics gate is open. The Monte Carlo code MCF, presented in this work, represents a new evaluation technique for the estimation of gate utilisation factors. It uses the die-away profile of a neutron coincidence chamber generated either by MCNP TM , or by other means, to simulate the neutron detection arrival time pattern originating from independent spontaneous fission events. A shift register simulation algorithm, embedded in the MCF code, then calculates the coincidence counts scored within the electronics gate. The gate utilisation factor is then deduced by dividing the coincidence counts obtained with that obtained in the same Monte Carlo run, but for an ideal detection system with a coincidence gate utilisation factor equal to unity. The MCF code has been benchmarked against analytical results calculated for both single and double exponential die-away profiles. These results are presented along with the development of the closed form algebraic expressions for the two cases. Results of this validity check showed very good agreement. On this basis, previously published analytical results for the double exponential case are thought to be in error. As derived analytically, the numerical calculations have been found to be both independent of the detector's efficiency and of the spontaneous fission neutron multiplicity distribution used in the Monte Carlo calculations. Extension of the MCF calculations to multiplicity counting, and in particular to triple coincidence counting, confirmed that, for a single exponential die-away profile, the triple gate utilisation factor is equal to the square of the real gate utilisation factor. For other profiles this relation no longer holds. An analytical expression is given for the case of a double exponential profile. Comparison of the MCF results with earlier calculated estimates of the gate utilisation factor for the on-site laboratory neutron coincidence chamber showed a significant difference. Use of the MCF results led to much better agreement between the observed and calculated specific reals coincidence rate of the on-site laboratory counter for the assay of plutonium samples. Moreover, the present work constitutes a further step towards the improvement of the accuracy of absolute Monte Carlo calculations for active or passive neutron measurements of nuclear materials.

MCNP modelling of a combined neutron/gamma counter

A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, ε, to, α, the ratio of (α,n) to spontaneous fission neutron emission rate and to f R, the reals coincidence gate utilisation factor, are presented. Sources of the inaccuracy in the calculations have not yet been fully investigated, because of the vast parameter space to be considered, but values of the coincidence gate utilisation factor derived directly from the MCNP data have been found to be overestimated by about 8.2%. Once bias-corrected, the trends of the real coincidence counts rate as a function of sample mass for three types of sample could be matched to experimental results within 0.33%. This result confirms the possible use of MCNP to calculate response trends accurately for a wide variety of source materials, given a limited experimental calibration set.

Radiometric measurements on non-destructive assay standards fabrication for WIPP Performance Demonstration Program

The Inorganic Elemental Analysis Group of LANL has prepared several different sets of working reference materials (WRMs). These WRMs are prepared by blending quantities of nuclear materials (plutonium, americium, and enriched uranium) with diatomaceous earth. The blends are encapsulated in stainless steel cylinders. These WRMs are being measured as blind controls in neutron and gamma based non-destructive assay (NDA) instruments. Radiometric measurements on the blending homogeneity and verification on a set of sixty-three plutonium based WRMs are discussed in this paper.

Integration of NDA Instruments into the Savannah River Plant Computer Network

The Safeguards Systems Group at the Los Alamos National Laboratory is collaborating with several other national laboratories in designing an integrated, state-of-the-art nondestructive assay (NDA) system for a new facility being constructed at the Savannah River Plant. We have focused our efforts on the development of an Instrument Control Computer (ICC) that will perform interfacing and communication tasks among small computers controlling NDA instruments in the Sample Assay Room and the Feed Assay Room, the host computer for the process control system, and an existing nuclear materials accounting computer. Solution sample scheduling, feed-material sample verification, and other ICC functions will be described in detail.

Nondestructive Assay System Development for a Plutonium Scrap Recovery Facility

A plutonium scrap recovery facility is being constructed at the Savannah River Plant (SRP). The safeguards groups of the Los Alamos National Laboratory have been working since the early design stage of the facility with SRP and other national laboratories to develop a state-of-the-art assay system for this new facility. Not only will the most current assay techniques be incorporated into the system, but also the various nondestructive assay (NDA) instruments are to be integrated with an Instrument Control Computer (lCC). This undertaking is both challenging and ambitious; an entire assay system of this type has never been done before in a working facility. This paper will describe, in particular, the effort of the Los Alamos Safeguards Assay Group in this endeavor. Our effort in this project can be roughly divided into three phases: NDA development, system integration, and integral testing.

Nondestructive Assay Instrumentation for a Savannah River Plant Upgrade Project

We have designed and are developing three different computer-based spectrometer systems. Two will measure the concentration of Pu solutions by gamma-ray and by stimulated x-ray fluorescence emissions. The third will measure relative isotopic abundances from the gamma-ray emissions of solid samples in closed containers. All systems are coupled to remote terminals and bar code readers, and also to mini-computer based MCAs, which in-turn are linked to another computer to provide a state-of-the-art NDA measurement capability. Installation at SRP is planned in late 1985.