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.

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Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

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.

/pdf/pwr-and-bwr-spent-fuel-assembly-gamma-spectra-measurements-3hcovl2isy.pdf

PWR and BWR spent fuel assembly gamma spectra measurements

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is “technology driven” in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

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

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

Determining plutonium in spent fuel with nondestructive assay techniques

3He replacement for nuclear safeguards applications - an integrated test program to compare alternative neutron detectors

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.

Experimental and Analytical Plans for the Non-Destructive Assay System of the Swedish Encapsulation and Repository Facilities

A Monte Carlo based spent fuel analysis safeguards strategy assessment

This paper describes an unattended mode neutron measurement that can provide the enrichment of the uranium in UF{sub 6} cylinders. The new passive neutron measurement provides better penetration into the uranium mass than prior gamma-ray enrichment measurement methods. The Passive Neutron Enrichment Monitor (PNEM) provides a new measurement technique that uses passive neutron totals and coincidence counting together with neutron self-interrogation to measure the enrichment in the cylinders. The measurement uses the neutron rates from two detector pods. One of the pods has a bare polyethylene surface next to the cylinder and the other polyethylene surface is covered with Cd to prevent thermal neutrons from returning to the cylinder. The primary neutron source from the enriched UF{sub 6} is the alpha-particle decay from the {sub 234}U that interacts with the fluorine to produce random neutrons. The singles neutron counting rate is dominated by the {sub 234}U neutrons with a minor contribution from the induced fissions in the {sub 235}U. However, the doubles counting rate comes primarily from the induced fissions (i.e., multiplication) in the {sub 235}U in enriched uranium. The PNEM concept makes use of the passive neutrons that are initially produced from the {sub 234}U reactions that track the more » {sub 235}U enrichment during the enrichment process. The induced fission reactions from the thermal-neutron albedo are all from the {sub 235}U and provide a measurement of the {sub 235}U. The Cd ratio has the desirable feature that all of the thermal-neutron-induced fissions in {sub 235}U are independent of the original neutron source. Thus, the ratio is independent of the uranium age, purity, and prior reactor history. « less

Howard O Menlove

Papers

Determining plutonium in spent fuel with nondestructive assay techniques

3He replacement for nuclear safeguards applications - an integrated test program to compare alternative neutron detectors

Experimental and Analytical Plans for the Non-Destructive Assay System of the Swedish Encapsulation and Repository Facilities

A Monte Carlo based spent fuel analysis safeguards strategy assessment

A more accurate and penetrating method to measure the enrichment and mass of UF6 storage containers using passive neutron self-interrogation