[en] Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties

[en] Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of ²⁴¹Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased (α,n) production in oxide fuels from the ²⁴¹Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of ²⁴¹Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of ²⁴¹Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of ²⁴¹Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

[en] The integration of Self-Interrogation Neutron Resonance Densitometry (SINRD) and Differential Die-away Self-Interrogation (DDSI) to more accurately quantify plutonium mass in a PWR 17x17 spent fuel assembly in water has been investigated via Monte Carlo N-Particle eXtended transport code simulations. Both of these instruments utilize the ²⁴⁴Cm spontaneous fission neutrons to self-interrogate the fuel pins. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. Thus, the ²³⁵U and ²³⁹Pu content in the spent fuel can be measured using ²³⁵U and ²³⁹Pu fission chambers placed adjacent to the assembly. DDSI uses time correlation neutron counting in a ³He-based detector system, with triggering on the spontaneous fission events, to determine the spontaneous fission rate and the induced fission rate in the spent fuel. The time separation of counts from spontaneous fission (early gate) and induced fission (late gate) neutrons enables DDSI to independently measure the fertile mass and fissile mass in spent fuel. Based on the complementary characteristics of SINRD and DDSI, we believe that the integration of these instruments will provide a more robust verification method for spent fuel assemblies by improving the ability to measure plutonium. This research is part of the Next Generation Safeguards Initiative (NGSI) of the U.S. DOE and is being conducted in collaboration with several other national laboratories and universities throughout the U.S. Future work includes performing experimental measurements with both SINRD and DDSI on LWR fresh and spent fuel in water. (author)

[en] In case of the Pu mass determination in scattered powder in a glovebox using Continuous Neutron Monitor (CNM) with totals, self-multiplication of neutron (M) and alpha value are properly required to be set. M of scattered powder can be easily estimated by a simulation code, but it is very difficult to estimate alpha value by such a simulation because interactions between alpha ray generated from plutonium and impurities in the powder are not consistent. Therefore, we tried to examine an estimation technique of alpha value by direct measurement. As a result, by measuring the samples taken from the scattered powder using a multiplicity counter with a dual ring structure of ³He tubes, we could confirm a good correlation between ring ratio (inner ring count rates / outer ring count rates) and the alpha values. Thus, we can estimate alpha value in the powder directly by the ring ratio measurement. By applying this technique to CNM and designing a new detector with a double layer structure of neutron detection tubes, we had a prospect that CNM would be able to measure the plutonium mass continuously. (author)

[en] The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the ²³⁵U and ²³⁹Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the ²⁴⁴Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using ²³⁵U and ²³⁹Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

[en] In order to implement facility nuclear material accountancy (NMA) and safeguards properly, it is important to understand where and how much holdup deposit in the process is present for the planning of the cleanout before PIT. JAEA and LANL developed and implemented a Glovebox Cleanout Assistance Tool (BCAT) to assist cleanout activity (Recovering MOX powder in a glovebox) for invisible holdup effectively by computational approach which is called distributed source-term analysis (DSTA). To know the holdup location and activity, the BCAT tool uses a set of simple neutron slab detectors and a matrix showing the relation between source activity and measured intensity, The matrix is defined by the MCNPX simulation using known source activity (be check source) at 53 source voxels and measured intensity at 57 measurement positions almost uniformly distributed in a process room. The model of MCNPX for entire process is very precisely established. We have implemented and experimentally proved that the BCAT tool can direct the operator to recoverable holdup that would otherwise be accounted for as MUF. Reducing facility MUF results in a direct improvement of the facility NMA. It is expected that using the BCAT tool over time will reduce the holdup in the conversion room and the total MUF declaration at the facility, as well as enable the staff to significantly improve their knowledge of the locations of residual holdup in the process area. JAEA would like to use this application for dismantling of the glovebox with transparency in the future. (authors)

[en] Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the ²³⁹Pu and ²³⁵U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of ²³⁹Pu and ²³⁵U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that ²³⁵U content has a major impact on the SINRD signal in addition to the ²³⁹Pu content. Plutonium-241 and ²⁴¹Am are the two main nuclides responsible for the variation in the count rate ratio with cooling time. In short, this work provides insights into some of the main factors that affect the performance of SINRD, and it should help improve the hardware design and the algorithm used to interpret the signal for the SINRD technique. In addition, the modeling and simulation techniques used in this work can be easily adopted for analysis of other NDA systems, especially when complex systems like spent nuclear fuel are involved. These studies were conducted at Los Alamos National Laboratory

[en] Prototype safeguards instrument for nuclear material accountancy (NMA) of uranium/transuranic (U/TRU) products that could be produced in a future advanced PWR fuel processing facility has been developed and characterized. This is a new, hybrid neutron measurement system based on fast neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) methods. The FNEM method is sensitive to the induced fission rate by fast neutrons, while the PNAR method is sensitive to the induced fission rate by thermal neutrons in the sample to be measured. The induced fission rate is proportional to the total amount of fissile material, especially plutonium (Pu), in the U/TRU product; hence, the Pu amount can be calibrated as a function of the induced fission rate, which can be measured using either the FNEM or PNAR method. In the present study, the prototype system was built using six "3He tubes, and its performance was evaluated for various detector parameters including high-voltage (HV) plateau, efficiency profiles, dead time, and stability. The system's capability to measure the difference in the average neutron energy for the FNEM signature also was evaluated, using AmLi, PuBe, "2"5"2Cf, as well as four Pu-oxide sources each with a different impurity (Al, F, Mg, and B) and producing (α,n) neutrons with different average energies. Future work will measure the hybrid signature (i.e., FNEM×PNAR) for a Pu source with an external interrogating neutron source after enlarging the cavity size of the prototype system to accommodate a large-size Pu source (~600 g Pu). - Highlights: • Prototype safeguards instrument for nuclear material accountancy has been developed and characterized. • The prototype system is based on a hybrid measurement technique (FNEM and PNAR). • Various detector parameters (i.e., efficiency profile, dead time, and stability) were evaluated. • The system's capability to measure the difference in the average neutron energy for the FNEM signature was evaluated.

[en] Correlated neutron counting using multiplicity shift register logic extracts the first three factorial moments from the detected neutron pulse train. The descriptive properties of the measurement item (mass, the ratio of (α,n) to spontaneous fission neutron production, and leakage self-multiplication) are related to the observed singles (S), doubles (D) and triples (T) rates, and this is the basis of the widely used multiplicity counting assay method. The factorial moments required to interpret and invert the measurement data in the framework of the point kinetics model may be calculated from the spontaneous fission prompt neutron multiplicity distribution P(ν). In the case of "2"3"8U very few measurements of P(ν) are available and the derived values, especially for the higher factorial moments, are not known with high accuracy. In this work, we report the measurement of the triples rate per gram of "2"3"8U based on the analysis of a set of measurements in which a collection of 10 cylinders of UO_2F_2, each containing about 230 g of compound, were measured individually and in groups. Special care was taken to understand and compensate the recorded multiplicity histograms for the effect of random cosmic-ray induced background neutrons, which, because they also come in bursts and mimic fissions but with a different and harder multiplicity distribution. We compare our fully corrected (deadtime, background, efficiency, multiplication) experimental results with first principles expectations based on evaluated nuclear data. Based on our results we suspect that the current evaluated nuclear data is biased, which points to a need to undertake new basic measurements of the "2"3"8U prompt neutron multiplicity distribution

[en] Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio was more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.