Vladimir Henzl

Integrated Nondestructive Assay Systems to Estimate Plutonium in Spent Fuel Assemblies

Abstract An integrated nondestructive assay (NDA) system combining active (neutron generator) and passive neutron detection and passive gamma (PG) detection is being analyzed in order to estimate the amount of plutonium, verify initial enrichment, burnup, and cooling time, and detect partial defects in a spent fuel assembly (SFA). Active signals are measured using the differential die-away (DDA), delayed neutron (DN), and delayed gamma (DG) techniques. Passive signals are measured using total neutron (TN) counts and both gross and spectral resolved gamma counts. To quantify how a system of several NDA techniques is expected to perform, all of the relevant NDA techniques listed above were simulated as a function of various reactor conditions such as initial enrichment, burnup, cooling time, assembly shuffling pattern, reactor operating conditions (including temperature, pressure, and the presence of burnable poisons) by simulating the NDA response for five sets of light water reactor assemblies. This paper compares the performance of several exploratory model-fitting options (including neural networks, adaptive regression with splines, iterative bias reduction smoothing, projection pursuit regression, and regression with quadratic terms and interaction terms) to relate data simulated with measurement and model error effects from various subsets of the NDA techniques to the total Pu mass. Isotope masses for SFAs and expected detector responses (DRs) for several NDA techniques are simulated using MCNP, and the DRs become inputs to the fitting process. Such responses include eight signals from DDA, one from DN, one from TN, and up to seven from PG; the DG signal will be examined separately. Results are summarized using the root-mean-squared estimation error for plutonium mass in held-out subsets of the data for a range of model and measurement error variances. Different simulation assumptions lead to different spent fuel libraries relating DRs to Pu mass. Some results for training with one library and testing with another library are also given.

A priori precision estimation for neutron Triples counting

The nondestructive assay of Plutonium bearing items for criticality, safety, security, safeguards, inventory balance, process control, waste management and compliance is often undertaken using correlated neutron counting. In particular Multiplicity Shift Register analysis allows one to extract autocorrelation parameters from the pulse train which can, within the framework of a simple interpretational model, be related to the effective 240Pu spontaneous fission mass present. The effective 240Pu mass is a weighted sum of the 238Pu, 240Pu and 242Pu masses so if the relative isotopic composition of the Pu can be established from the measured 240Pu effective mass one can estimate the total Pu mass and also the masses of the individual isotopes, example the fissile species 239Pu and 241Pu. In multiplicity counting three counting rates are obtained. These are the Singles, Doubles and Triples rates. The Singles rate is just the gross, totals or trigger rate. The Doubles and Triples rates are calculated from factorial moments of the observed signal triggered neutron multiplicity distributions following spontaneous fission in the item and can be thought of as the rate of observed coincident pairs and coincident triplets on the pulse train. Coincident events come about because the spontaneous fission and induced fission chains taking place in the item result in bursts of neutrons. These remain time correlated during the detection process and so retain information, through the burst size distribution, about the Pu content. In designing and assessing the performance of a detector system to meet a given goal it is necessary to make a priori estimates of the counting precision for all three kinds of rates. This is non-trivial because the counting does not obey the familiar rules of a Poissonian counting experiment because the pulse train has time correlated events on it and the train is sampled by event triggered gates that may overlap. For Singles and Doubles simple approximate analytical empirical rules for how to estimate the variance have been developed guided by theory and refined by experiment. However, for Triples no equivalent rules have been put forward and tested until now. In this work we propose an analytical expression, the CSH relation, for the variance on the Triples count and exercise it against experimental data gathered for Pu items measured in the Los Alamos National Laboratorys Epithermal Neutron Multiplicity Counter (ENMC). Preliminary results are encouraging and reasonable agreement with observation, considered fit for scoping studies, is obtained. We have also looked at the behavior using Monte Carlo simulations.

Measurements of the neutron-proton and neutron-carbon total cross section from 150 to 800 keV

There have been very few measurements of the total cross section for np scattering below 500 keV. In order to differentiate among NN potential models, improved cross section data between 20 and 600 keV are required. We measured the np and nC total cross sections in this energy region by transmission; a collimated neutron beam was passed through CH2 and C samples and transmitted neutrons were detected by a BC-501A liquid scintillator. Cross sections were obtained with a precision of 1.1-2.0% between 150 and 800 keV using ratios of normalized neutron yields measured with and without the scattering samples in the beam. In energy regions where they overlap, the present results are consistent with existing precision measurements, and fill in a significant gap in the data between En = 150 and 500 keV.

Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238U first fission contribution was found to not affect the DDA performance in reconstructing 239Pueff. As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

Uncertainty Quantification for New Approaches to Spent Fuel Assay

Abstract Estimating plutonium (Pu) mass in spent nuclear fuel assemblies (SFAs) helps inspectors ensure that no Pu is diverted. Therefore, nondestructive assay (NDA) methods are being developed to assay Pu mass in SFAs. Uncertainty quantification is an important task in most assay methods, and particularly for SFA assay. A computer model (MCNPX) is being used to predict isotope masses and the spatial distribution of masses in virtual SFAs for 64 combinations of initial fuel enrichment (IE), fuel utilization [burnup (BU)], and cooling time (CT) values. Additional MCNPX modeling for the same 64 virtual SFAs provided the expected detector responses (DRs) for several NDA techniques such as the passive neutron albedo reactivity method and the 252Cf interrogation with prompt neutrons method. A previous paper describes one uncertainty quantification approach involving Monte Carlo (MC) simulation using individually any of six new NDA options together with IE, BU, and CT. This paper provides an interpretation of the MC approach that is suited for a numerical Bayesian alternative, separately assesses the impact of MCNPX interpolation error, and compares several options to use subsets of IE, BU, CT, and one DR.

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.

Differential Die-Away Instrument: Report on Benchmark Measurements and Comparison with Simulation for the Effects of Neutron Poisons

In this report, new experimental data and MCNPX simulation results of the differential die-away (DDA) instrument response to the presence of neutron absorbers are evaluated. In our previous fresh nuclear fuel experiments and simulations, no neutron absorbers or poisons were included in the fuel definition. These new results showcase the capability of the DDA instrument to acquire data from a system that better mimics spent nuclear fuel.

Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDA instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.

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 3He 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.

Differential Die-Away Instrument: Report on Neutron Detector Recovery Performance and Proposed Improvements

Four helium-3 (3He) detector/preamplifier packages (¾”/KM200, DDSI/PDT-A111, DDA/PDT-A111, and DDA/PDT10A) were experimentally tested to determine the deadtime effects at different DT neutron generator output settings. At very high count rates, the ¾”/KM200 package performed best. At high count rates, the ¾”/KM200 and the DDSI/PDT-A111 packages performed very well, with the DDSI/PDT-A111 operating with slightly higher efficiency. All of the packages performed similarly at mid to low count rates. Proposed improvements include using a fast recovery LANL-made dual channel preamplifier, testing smaller diameter 3He tubes, and further investigating quench gases.