Applications of Nuclear Techniques (CRETE13) International Journal of Modern Physics: Conference Series Vol. 27 (2014) 1460130 (8 pages) The Authors DOI: 10.1142/S2010194514601306 FISSILE MATERIALS DETECTION VIA NEUTRON DIFFERENTIAL DIE-AWAY TECHNIQUE V. F. BATYAEV, O. V. BOCHKAREV, S. V. SKLYAROV All-Russia Research Institute of Automatics (VNIIA), Moscow 127055, Russia Published 25 February 2014 This work is devoted to the differential die-away technique that is widely used for active detection of fissile materials via pulsed neutron generators. The technique allows direct detection of milligram quantities of uranium-235 and plutonium-239 in objects with volumes up to several cubic meters. Our group has demonstrated this technique, creating a special installation based on the commercially produced ING-07T pulsed neutron generator. The installation includes eight proportional 3 Не-counters mounted inside a polyethylene moderator with a cadmium filter, as well as a polyethylene chamber into which a 70-liter container is loaded for inspection. Preliminary testing showed that the minimum detectable mass of unshielded uranium-235 is ~3 mg, using a 5 10 8 n/s neutron yield and 8 min measurement time. When the container is filled with neutron absorbing materials, e.g., iron, the minimum detectable mass increases to ~30 mg. Use of borated screens further increases the minimum mass that can be detected. The tested installation and/or its modifications can be used for control and detection of fissile materials in various applications from luggage inspection to control containers with nuclear fuel cycle radioactive wastes. Keywords: Differential die-away technique; neutron generator; 235 U and 239 Pu content in closed waste containers. 1. Introduction Measuring the amount of special nuclear material (SNM) is an important challenge, needed for material control and accountability purposes, radiation safety and criticality issues at nuclear installations, radioactive waste management sites, and for the containment and security of SNM. The mission to control SNM in waste containers from nuclear production is quite specific. It is complicated to implement non-destructive assay in practice, due to the presence of matrix materials with unknown type and density in sealed containers. Passive gamma-spectrometric measurements can have large errors due to shielding of low-energy gamma-rays emitted naturally from 235 U and 239 Pu in SNM and matrix materials. For this problem, it is appropriate to use active control techniques with external neutron sources that produce more penetrating radiation compared to This is an Open Access article published by World Scientific Publishing Company. It is distributed under the terms of the Creative Commons Attribution 3.0 (CC-BY) License. Further distribution of this work is permitted, provided the original work is properly cited. 1460130-1
V. F. Batyaev, O. V. Bochkarev & S. V. Sklyarov naturally emitted gamma-rays. This reduces the issue of self-shielding and results in more informative and sensitive systems for the control of SNM, such as 235 U and 239 Pu. 2. SNM Detection Technique and its Design Realization at VNIIA The base configuration of the nuclear waste control set-up was fabricated at VNIIA in 2011. This set-up consists of a hollow polyethylene cube with built-in 3 He counters located inside cadmium screens (Fig. 1). The set-up configuration corresponds to the fission neutron measurement technique known as differential die-away. 1-2 When a pulsed neutron source is used, the time distribution of rates in the 3 He counters is described by: 3 Ф(t) = А ехр(-αt) + B ехр(-βt) + C, (1) where t is the time after the fast neutron pulse. The first term in Eq. (1) corresponds to neutrons emitted from the neutron source and partly to the epicadmium neutrons, while the second term corresponds to the 235 U fissions induced by thermal neutrons. The third term C arises from the sum of responses from delayed neutrons, spontaneous fission neutrons of various nuclides and neutrons produced in (,n) reactions. The parameters and are time constants of thermal neutron die-away rates, which depend on properties of the moderator, inside and outside of the cadmium screen, respectively. For the parameter, such properties include polyethylene dimensions and geometry, as well as the physics of neutron interactions in the container and the materials inside. Selection of these parameters (described in our previous simulation studies 3 ) allowed us to achieve reliably distinguishable exponents in Eq. (1) and optimal sensitivity of to the type and quantity of material inside the container. D-T Horizontal cross section Container 235 U with matrix D-T Vertical cross section lid Container 235 U with matrix bottom Polyethylene 3 He-counters inside polyethylene assemblies covered by Cd (8 units) 160 liter cavity 45x45x80cm 70 liter container 35x35x60cm Fig. 1. Chart of nuclear waste control set-up. The fabricated set-up uses the VNIIA-produced ING-07Т pulsed neutron generator that provides 14 MeV neutrons with maximum neutron yield of ~10 9 n/s, pulse frequency adjustable within the range of 0.4 to 10 khz and pulse duration from 20 to 100 µs. The 1460130-2
Fissile Materials Detection via Neutron Differential Die-away design of the system also allows installation of an alternative neutron pulse generator with higher neutron yield. A 70 liter steel container with matrix simulants and uranium samples is loaded into the chamber from the top through a removable polyethylene lid. Signals from the 3 Не counters are processed with a 128-channel time analyzer connected to a remote laptop. Fig. 2. System for active neutron measurement of SNM amount in waste of nuclear production. 3. Experiments and Data Analysis To model the matrix elements of a container, graphite bars, steel L-bars and polyethylene plates of various sizes were fabricated and placed inside the container. Moreover, the central part of the AT-400R container with a boron nitride hollow sphere inside was used to simulate a highly-absorbing neutron material. To model SNM, depleted and enriched uranium samples were used with a 235 U mass of 0.35 to 26 g. In total, we performed several dozen measurements of the time distribution of the 3 He counter rates for cases of different mass SNM samples placed inside the container filled with different matrix simulants. Each measurement lasted 8 to 15 minutes and used 5 10 8 n/s yield of 14 MeV neutrons from the ING-07T generator operating on 400 Hz frequency. Figure 3 shows some examples of measured rate distributions as a function of time after the neutron pulse. The ratio of fission neutron response to source neutron response can be an informative parameter, since it depends on amount of 235 U present, but does not depend on the neutron generator yield. The ratio of the amplitudes B/A in the two exponential terms in Eq. (1) can be used as such a parameter. The measured relationships of В/А to the amount of 235 U can be regarded as calibration curves of the given set-up for different matrix types and densities and so can be further used to determine the 235 U amount in controlled containers. A set of such calibration curves is shown in Fig. 4. 1460130-3
V. F. Batyaev, O. V. Bochkarev & S. V. Sklyarov Fig. 3. Time distributions of fission neutrons responses for 235 U samples in an empty container (no matrix) for an 8 minute measurement time and 5 10 8 n/s neutron yield. Fig. 4. Calibration relationship of В/А to mass of 235 U located in the container center. Extrapolating the measured experimental data towards the lower values, the minimum detectable SNM mass for various matrix materials in the container can be identified. Figure 5 illustrates such extrapolation towards the double error values for the respective background measurements (based on the Neyman Pearson lemma 4 ). The minimum detectable masses obtained in such a way correspond to a 0.84 detection probability and a 0.16 probability of false alarm. To get higher detection probabilities, one needs to use higher quantiles of background uncertainty providing thereby higher values of minimum 1460130-4
Fissile Materials Detection via Neutron Differential Die-away detectable masses. For instance, the minimum detectable masses would be 1.7 and 2.6 times higher in the case of 0.95 and 0.99 detection probabilities, respectively. A unified power function of the form B/A=k m 0.9, where m is mass, k is a proportionality coefficient for each matrix, was selected to approximate the curves for the experimental data. The results of this procedure are presented in Table 1. B/A - B/A (background), 10-2 1 0.1 0.01 1E-3 Graphite 0.62g/cm 3 Graphite 0.31g/cm 3 No matrix Fe 0.95g/cm 3 Fe 1.59g/cm 3 Depleted U in Boron sphere (C/A) 2σ of backgrounds 0.01 0.1 1 10 235 U mass, g Fig. 5. Extrapolation of experimental data to assess the minimum detectable SNM mass for various matrix materials in container. The data obtained with the 0.16 g/cm 3 polyethylene matrix are not shown since they are close to the 0.62 g/cm 3 graphite data. Table 1. Minimum detectable mass of SNM (extrapolation). # Matrix type 235 U minimum detectable mass (mg) 1 No matrix 4.3 ± 1.5 2 Polyethylene 0.16g/cm 3 2.1 ± 0.8 3 Graphite 0.31 g/cm 3 2.7 ± 0.9 4 Graphite 0.62 g/cm 3 2.3 ± 0.8 5 Iron 0.95 g/cm 3 25 ± 9 6 Iron 1.59 g/cm 3 32 ± 11 7 AT-400R container 75 ± 25 (in depleted uranium) 0.1 0.01 1E-3 1E-4 C/A, 10-2 As the result of further processing of the experimental data, the information parameter В/А was determined as a function of not only the SNM mass but also the SNM location inside the container. In particular, В/А was measured when a 235 U sample moves inside the container vertically (Fig. 6), along (Fig. 7) and across (Fig. 8) the ING-07 axis. 1460130-5
V. F. Batyaev, O. V. Bochkarev & S. V. Sklyarov Fig. 6. В/А parameter vs. position of SNM along the height of the container for different matrices. Fig.7. Function of В/А vs. SNM position along the ING07 axis. The distributions presented in Figs. 6 to 8 are determined by: Limited length of 3 He counters and physical features of the matrix materials used in the case of the vertical distribution (Fig. 6); Dependence of the thermal neutron field center location on the type of matrix in the case of the longitudinal distribution (Fig. 7); and Symmetry of counter locations over the longitudinal axis, thereby providing a practically stable lateral distribution (Fig. 8) of summed rates of 3 He counters. 1460130-6
Fissile Materials Detection via Neutron Differential Die-away Fig. 8. Function of В/А vs. SNM position across the ING07 axis (no matrix in the container). The observed longitudinal and lateral distributions (Figs. 7 and 8) prove a possibility of using individual rates of 3 He counters to determine the horizontal coordinates of SNM location in a container. Determination of the SNM vertical position is not possible without having 3 He counters in the bottom and the lid of the set-up. The experiments have also indicated that the time constant β of the thermal neutron fission response die-away rate depends only on the type and density of matrix and is almost constant over the range of measured uranium mass (Fig. 9). This can be used to determine the type of matrix without additional measurements and apply the appropriate calibration dependence to measure SNM mass. Fig. 9. Function of time constant β vs. SNM mass and various matrices. Lines indicate the averaged values for this type of matrix. 1460130-7
V. F. Batyaev, O. V. Bochkarev & S. V. Sklyarov 4. Conclusion The technique and experimental set-up for active neutron measurement of the amount of SNM in medium size containers have been tested. Use of a pulsed neutron generator combined with the differential neutron die-away technique allows high sensitivity detection of SNM responses from nuclear production waste in containers without opening them. In the absence of neutron-absorbing materials in a container, the minimum detectable SNM mass is 0.002 0.004 g of 235 U for measurement times of 8 to 15 min. In the case where such materials are present, the minimum detectable mass can be higher by an order of magnitude or more. We presented results only for 235 U because we had no possibility to measure 239 Pu in our lab. However, we consider that a 239 Pu measurement would be more accurate, providing thereby a lower minimum detectable mass due to the higher thermal neutron fission cross-sections of 239 Pu as compared to 235 U. The experiments prove that it is possible to identify the type of matrix without opening a container. The type and density of matrix material are determined by the time constant β of the thermal neutron die-away rate in a container, and the constant itself is almost unaffected by SNM mass. In turn, determining the matrix type and density allows one to choose the appropriate calibration curve of counting rate versus amount of 235 U and to reach reliable accuracies of measuring SNM in containers. The novelty of our work is that we can detect milligram quantities of 235 U using only eight 3 He counters. Most similar assay systems use many more counters (see Ref. 5, for example). In addition, we have obtained longitudinal and lateral distributions to prove it is possible to use the individual rates of 3 He counters to determine the horizontal coordinates of the location of SNM in a container. The results obtained can be used for designing a bigger experimental system (e.g., for assay of 200 liter drums containing nuclear production waste), as well as for measuring the remaining quantity of SNM in spent fuel assemblies and other types of containers where materials fissionable with thermal neutrons are or may be present. References 1. W. Rotter, Annals of Nuclear Science and Engineering 1(7 8), 451 (1974). 2. Kelly A. Jordan and Tsahi Gozani, Nucl. Instr. Meth. Phys. Res. B 261, 365 (2007). 3. Yury N. Barmakov, Evgeny P. Bogolyubov, Oleg V. Bochkarev, Yury G. Polkanov, Vadim L. Romodanov and Dina N. Chernikova, Int. J. Nuclear Energy Science and Technology 6(2), 127 (2011). 4. Yu. V. Prokhorov, ed., Probability and Mathematical Statistics Encyclopedia ( Great Russian Encyclopedia Publishing House, 1999). 5. A.-C. Raoux, J. Loridon, A. Mariani and C. Passard, Nucl. Instr. Meth. Phys. Res. B 266, 4837 (2008). 1460130-8