Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

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 137 Cs, 154 Eu, and 134 Cs 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.

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PWR and BWR spent fuel assembly gamma spectra measurements

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 Laboratory's 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.

A priori precision estimation for neutron Triples counting

Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel. (August 2011) Adrienne Marie LaFleur, B.S., Texas A&M University Chair of Advisory Committee: Dr. William S. Charlton The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: 1) SINRD provides absolute measurements of burnup independent of the operator’s declaration.

Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel

Verification of fuel parameters is a central undertaking for nuclear inspectors aiming at verifying the completeness and correctness of operator declarations. Traditionally, such verification is do ...

Determination of spent nuclear fuel parameters using modelled signatures from non-destructive assay and Random Forest regression

Ever since there has been spent fuel (SF), researchers have made nondestructive assay (NDA) measurements of that fuel to learn about its content. In general these measurements have focused on the simplest signatures (passive photon and total neutron emission) and the analysis has often focused on diversion detection and on determining properties such as burnup (BU) and cooling time (CT). Because of shortcomings in current analysis methods, inspectorates and policy makers are interested in improving the state-of-the-art in SF NDA. For this reason the U.S. Department of Energy, through the Next Generation Safeguards Initiative (NGSI), targeted the determination of elemental Pu mass in SF as a technical goal. As part of this research effort, 14 nondestructive assay techniques were studied . This wide range of techniques was selected to allow flexibility for the various needs of the safeguards inspectorates and to prepare for the likely integration of one or more techniques having complementary features. In the course of researching this broad range of NDA techniques, several cross-cutting issues were. This paper will describe some common issues and insights. In particular we will describe the following: (1) the role of neutron absorbers with emphasis on how these absorbers vary in SFmore » as a function of initial enrichment, BU and CT; (2) the need to partition the measured signal among different isotopic sources; and (3) the importance of the “first generation” concept which indicates the spatial location from which the signal originates as well as the isotopic origins.« less

Technical Cross-Cutting Issues for the Next Generation Safeguards Initiative's Spent Fuel Nondestructive Assay Project

There are a variety of motivations for quantifying plutonium in used (spent) fuel assemblies by means of nondestructive assay including the following: shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories or fuel storage facilities. Twelve NDA techniques were identified that provide information about the composition of an assembly. Unfortunately, none of these techniques is capable of determining the Pu mass in an assembly on its own. However, it is expected that the Pu mass can be quantified by combining a few of the techniques. Determining which techniques to combine and estimating the expected performance of such a system is the purpose of the research effort recently begun. The research presented here is a complimentarily experimental effort. This paper will focus on experimental results of one of the twelve non-destructive assay techniques - passive neutron albedo reactivity. The passive neutron albedo reactivity techniques work by changing the multiplication the pin experiences between two separate measurements. Since a single spent fuel pin has very little multiplication, this is a challenging measurement situation for the technique. Singles and Doubles neutron count rate were measured at Oak Ridge National Laboratory for three different burnup pins to test the capability ofmore » the passive neutron albedo reactivity technique.« less

A. Rajasingam

Papers

Technical Cross-Cutting Issues for the Next Generation Safeguards Initiative's Spent Fuel Nondestructive Assay Project

Determining plutonium in spent fuel with nondestructive assay techniques