[en] Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

[en] An early component of the Joint Fuel Cycle Study (JFCS) between the United States and the Republic of Korea is a test of gram scale electrochemical recycling of spent fuel which is to be performed at Idaho National Laboratory (INL). Included in this test is the development of Nondestructive Assay (NDA) technologies that would be applicable for International Atomic Energy Agency (IAEA) safeguards of the electrochemical recycling process. Of upmost importance to safeguarding the fuel cycle associated with the electrochemical recycling process is the ability to safeguard the U/TRU ingots that will be produced in the process. For the gram scale test, the ingots that will be produced will have an expected thermal power of approximately 130 mW. To ascertain how well the calorimetric assay NDA technique can perform in assaying these ingots, Los Alamos National Laboratory (LANL) has characterized and calibrated a small solid-state calorimeter called the Small Sample Calorimeter (SSC3) to perform these measurements at LANL. To calibrate and characterize the SSC3, a series of measurements were performed using certified ²³⁸Pu heat standards whose power output is traceable back to the National Institute of Standards and Technology (NIST) electrical standards. The results of these measurements helped establish both the calibration of the calorimeter as well as the expected performance of the calorimeter in terms of its accuracy and precision as a function of thermal power of the item that is being measured. In this report, we will describe the measurements that were performed and provide a discussion of the results of these measurements.

[en] ²⁴¹Pu has the shortest half-life of the abundant plutonium isotopes present in reprocessed irradiated nuclear fuel with a value of approximately 14.3 years. It is important to know the half-life of ²⁴¹Pu with a higher fractional accuracy than that of the other plutonium isotopes because the half-life of ²⁴¹Pu and its associated uncertainty affects the estimation by decay calculation of both the total amount of separated plutonium in storage and the determination of the total plutonium mass by non-destructive assay. This paper addresses the determination of the ²⁴¹Pu half-life using nuclear calorimetry by the measurement of the thermal power as ²⁴¹Pu evolves in time from a sealed plutonium source, ideally initially rich in ²⁴¹Pu and chemically stripped of ²⁴¹Am. The absolute accuracy of nuclear calorimeters can be ensured over long periods of time (many years) using long-lived nuclear reference materials and/or traceable electrical heat standards. One can, therefore, expect nuclear calorimetry to offer an accurate way to determine the half-life of ²⁴1Pu, which is comparable in quality and independent, yet complementary, to other approaches. Temporal analysis of the power-versus-time data also yields an estimate of the specific power of ²⁴¹Pu, which other methods do not. After describing the principle of the method and developing the pertinent mathematical expressions, we outline the approach by drawing on some unpublished notes of Kenneth C. Jordan who carried out such experiments at the Mound Laboratory over 40 years ago. Today, Jordan’s work remains possibly the most significant experiment of its type to the ²⁴¹Pu nuclear data evaluator. However, objectively assigning confidence to his results is problematic because the details of the experiments and data reduction have never been adequately reported. This work goes some way to that end but, without the raw data and first-hand knowledge, cannot provide a complete record. We conclude that a new high-accuracy nuclear calorimetry campaign to re-measure the ²⁴¹Pu half-life and specific ²⁴¹Pu has the shortest half-life of the abundant plutonium isotopes present in reprocessed irradiated nuclear fuel with a value of approximately 14.3 years. It is important to know the half-life of ²⁴¹Pu with a higher fractional accuracy than that of the other plutonium isotopes because the half-life of ²⁴¹Pu and its associated uncertainty affects the estimation by decay calculation of both the total amount of separated plutonium in storage and the determination of the total plutonium mass by non-destructive assay. This paper addresses the determination of the ²⁴¹Pu half-life using nuclear calorimetry by the measurement of the thermal power as ²⁴¹Pu evolves in time from a sealed plutonium source, ideally initially rich in ²⁴¹Pu and chemically stripped of ²⁴¹Am. The absolute accuracy of nuclear calorimeters can be ensured over long periods of time (many years) using long-lived nuclear reference materials and/or traceable electrical heat standards. One can, therefore, expect nuclear calorimetry to offer an accurate way to determine the half-life of ²⁴¹Pu, which is comparable in quality and independent, yet complementary, to other approaches. Temporal analysis of the power-versus-time data also yields an estimate of the specific power of ²⁴¹Pu, which other methods do not. After describing the principle of the method and developing the pertinent mathematical expressions, we outline the approach by drawing on some unpublished notes of Kenneth C. Jordan who carried out such experiments at the Mound Laboratory over 40 years ago. Today, Jordan’s work remains possibly the most significant experiment of its type to the ²⁴¹Pu nuclear data evaluator. However, objectively assigning confidence to his results is problematic because the details of the experiments and data reduction have never been adequately reported. This work goes some way to that end but, without the raw data and first-hand knowledge, cannot provide a complete record. We conclude that a new high-accuracy nuclear calorimetry campaign to re-measure the ²⁴¹Pu half-life and specific

[en] Nondestructive Assay (NDA) techniques are used both by domestic inspectors and site level nuclear material management programs as well as by International Atomic Energy Agency inspectors to verify inventories of nuclear materials. The significant improvements in detector capabilities, electronics processing, and data analysis has led to new detection capabilities and greatly improved quantification of nuclear materials. Many of the improvements over the last decade have resulted from improved computing power. This has led to the ability to collect and analyze data in ways not possible only years ago. This paper will discuss some of the improvements of nondestructive assay technologies over the past several years and the implementation of these technologies in nuclear safeguards. (author)

[en] Nondestructive Assay (NDA) techniques are an important tool for the safeguarding of nuclear materials. NDA techniques are used by inspectors from both domestic agencies and international agencies such as the International Atomic Energy Agency as well as site level nuclear material management programs to verify that inventories of nuclear materials. This technology has been in development for over 40 years and significant improvements in detector capabilities, electronics processing and data analysis has lead to new detection capabilities and greatly improved quantification of nuclear materials. Many of the improvements over the last decade have resulted from improved computing power. This has lead to the ability to collect and analyze data in ways not possible only years ago. This poster will present some of the improvements of nondestructive assay technologies over the past several years and the implementation of these technologies in nuclear safeguards programs.

[en] The Dual Slab Verification Detector (DSVD) has been developed and built by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of ³He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. The DSVD will be used to perform measurements of the neutron flux emanating from inside the dry storage cask at several locations around each cask to establish a neutron 'fingerprint' that is sensitive to the contents of the cask. The sensitivity of the fingerprinting technique to the removal of specific amount of nuclear material from the cask is determined by the characteristics of the detector that is used to perform the measurements, the characteristics of the spent fuel being measured, and systematic uncertainties that are associated with the dry storage scenario. MCNPX calculations of the BN-350 dry storage asks and layout have shown that the neutron fingerprint verification technique using measurements from the DSVD would be sensitive to both the amount and location of material that is present within an individual cask. To confirm the performance of the neutron fingerprint technique in verifying the presence of BN-350 spent fuel in dry storage, an initial series of measurements have been performed to test the performance and characteristics of the DSVD. Results of these measurements will be presented and compared with MCNPX results.

[en] The attribute measurement technique provides a method for determining whether or not an item containing special nuclear material (SNM) possesses attributes that fall within an agreed upon range of values. One potential attribute is whether the mass of an SNM item is larger than some threshold value that has been negotiated as part of a nonproliferation treaty. While the historical focus on measuring mass attributes has been on using neutron measurements, calorimetry measurements may be a viable alternative for measuring mass attributes for plutonium-bearing items. Traditionally, calorimetry measurements have provided a highly precise and accurate determination of the thermal power that is being generated by an item. In order to achieve this high level of precision and accuracy, the item must reach thermal equilibrium inside the calorimeter prior to determining the thermal power of the item. Because the approach to thermal equilibrium is exponential in nature, a large portion of the time spent approaching equilibrium is spent with the measurement being within ∼10% of its final equilibrium value inside the calorimeter. Since a mass attribute measurement only needs to positively determine if the mass of a given SNM item is greater than a threshold value, performing a short calorimetry measurement to determine how the system is approaching thermal equilibrium may provide sufficient information to determine if an item has a larger mass than the agreed upon threshold. In previous research into a fast calorimetry attribute technique, a two-dimensional heat flow model of a calorimeter was used to investigate the possibility of determining a mass attribute for plutonium-bearing items using this technique. While the results of this study looked favorable for developing a fast calorimetry attribute technique, additional work was needed to determine the accuracy of the model used to make the calculations. In this paper, the results from the current work investigating the fast calorimetry attribute technique will be presented.

[en] Neutron emissions from fissioning nuclear material are temporally correlated. The detection of these correlated neutrons is frequently used to quantify plutonium (Pu) and other fissile materials for international nuclear safeguards and related activities. However, detector dead time affects the observed rates of correlated neutrons in a non-trivial manner, and must be accounted for to obtain accurate results. A major simplification made in the most widely used dead time corrections is that the neutron detections are occurring randomly in time. A few previous attempts at providing a dead time model for correlated neutrons have been limited in scope, have made simplifying assumptions early in the derivation, and have, in general, not been implemented in the broader safeguards community. This paper provides an exact dead time model for correlated neutron detections in a single channel system assuming an updating dead time, and therefore a paralyzable system. This dead time model includes the assumption that a single exponential, with one characteristic decay constant, can describe the system neutron die-away profile. An exact model for the effects of dead time on measured gate moments is derived which is extendable to an arbitrary order of neutron correlation. This dead time model predicts the measured gate moments based on the dead time and underlying detection rates, including the effects from detection rates with an arbitrarily high order of correlation. The effects of dead time on the apparent singles, doubles, triples and quadruples rates using either event triggered, random or mixed gate structure is also derived. Either the equations for the measured gate moments or the apparent multiplicity rates can be numerically inverted to find the dead time corrected multiplicity rates. Although the model has been explicitly solved for rates up to and including quadruples, it is directly extendable to any order of correlation. This model is presented from the perspective of neutron counting. However, the model is mathematically applicable to any series of events which are temporally correlated through an exponential probability distribution and which are subject to a dead time-like filter

[en] The Feynman-Y statistic is a type of autocorrelation analysis. It is defined as the excess variance-to-mean ratio, Y = VMR - 1, of the number count distribution formed by sampling a pulse train using a series of non-overlapping gates. It is a measure of the degree of correlation present on the pulse train with Y = 0 for Poisson data. In the context of neutron coincidence counting we show that the same information can be obtained from the accidentals histogram acquired using the multiplicity shift-register method, which is currently the common autocorrelation technique applied in nuclear safeguards. In the case of multiplicity shift register analysis however, overlapping gates, either triggered by the incoming pulse stream or by a periodic clock, are used. The overlap introduces additional covariance but does not alter the expectation values. In this paper we discuss, for a particular data set, the relative merit of the Feynman and shift-register methods in terms of both precision and dead time correction. Traditionally the Feynman approach is applied with a relatively long gate width compared to the dieaway time. The main reason for this is so that the gate utilization factor can be taken as unity rather than being treated as a system parameter to be determined at characterization/calibration. But because the random trigger interval gate utilization factor is slow to saturate this procedure requires a gate width many times the effective 1/e dieaway time. In the traditional approach this limits the number of gates that can be fitted into a given assay duration. We empirically show that much shorter gates, similar in width to those used in traditional shift register analysis can be used. Because the way in which the correlated information present on the pulse train is extracted is different for the moments based method of Feynman and the various shift register based approaches, the dead time losses are manifested differently for these two approaches. The resulting estimates for the dead time corrected first and second order reduced factorial moments should be independent of the method however and this allows the respective dead time formalism to be checked. We discuss how to make dead time corrections in both the shift register and the Feynman approaches.

[en] A novel class of scintillating materials, which are nanocomposites of known scintillators, is described. These nanocomposite materials are expected to have improved properties with respect to the properties of the bulk scintillators from which they are derived. Improvements include enhanced light output, reduced cost and greater size scalability. Optical properties of the nanocomposite phosphor components are described, with emphasis on the origin of the enhanced light yield. Finally, results from our first nanocomposite scintillator, including photopeak measurements, are described. The photopeak measurements demonstrate the proof-of-principle of the nanocomposite scintillator concept