Digital simulation of neutron and gamma measurement devices

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1 3Security of radioactive materials and transport 3 2 Digital simulation of neutron and gamma measurement devices A.-L. WEBER (IRSN) 1 - Maximum activity that a radioactive element can present without being detected. Context In application of the act of 25 July 1980 that provided for nuclear materials to be nationally controlled in France, IRSN carries out inspections covering the physical protection of nuclear materials, physical follow-up, and accountancy within French facilities. Some inspections give rise to physical verifications of nuclear materials held by operators, resorting to non-destructive measurement. The Institute s inspectors visit the nuclear fuel cycle facilities with their own, transportable devices to characterize the nuclear materials by making neutron counting and gamma-ray spectrometry measurements. The issue The main aims of developing modeling for the inspectors measuring devices are to optimize their performance and provide them with an additional calibration and assessment method in the event of discrepancies with the operators figures as stated in their local accountancy. In gamma spectrometry, modeling enables simulated spectra to be calculated for comparison against the experimental spectra. The PLUM (PLUtonium Masse) high-resolution experimental gamma-ray spectrometry system, used to quantify the plutonium present in waste drums is thus subject to modeling. The aim of this simulation study is to set up a database containing the detection 1 limits calculated in standard measurement configuration. In particular it enables us to assess the measuring time required. In neutron counting, digital simulation enables us to calculate neutron counts for comparison against experimental counts. The CHACAL passive neutron counter designed to quantify the plutonium present in elongated containers was PLUM experimental gamma-ray spectrometry system. modeled so that the deviations observed during inspection could be analyzed. The complexity of the geometry found in the nondestructive measuring devices prompted us to use a computation code based on the Monte- Carlo method. The code in question was MCNP (Monte-Carlo N-Particles), developed for calculating the transfer of neutrons, photons and electrons in three-dimensional geometry settings. Numerous individual histories were simulated from emission, through to their death by absorption or leakage into the medium crossed. The trajectory of a particle is thus broken down into sequences comprising free flight and at a given moment, a collision whose nature is randomly selected from a set of possible reactions in the material crossed, with probabilities linked to cross sections of the material in question. Thus a measuring device can be modeled and the expected response from the detection system (high-purity germanium detector in the case of gamma spectrometry, 3 He detectors in the case of neutron counting devices) calculated for a radioactive source, a container holding nuclear materials, or a drum containing neutron- and gamma-emitting waste. SCIENTIFIC AND TECHNICAL REPORT

code, enables us to estimate the spectrum of energy deposits in the detector for any gamma source facing it. Figure 2 Radiography of the crystal.

2 Figure 1 PLUM device. Digital simulation applied to the PLUM experimental gamma-ray spectrometry system Digital simulation of photon transfer in the germanium detector fitted to the PLUM device, using the MCNP computation code, enables us to estimate the spectrum of energy deposits in the detector for any gamma source facing it. Figure 2 Radiography of the crystal. THE PLUM DEVICE (PLUTONIUM MASS) The PLUM experimental gamma spectrometer (figure 1) is used to quantify plutonium masses present in technological waste drums ranging from a few milligrams to several tens of grams. It comprises a measurement bench and instrumentation in the form of a highly efficient high-purity coaxial germanium detector, a multichannel analyzer for spectra acquisition on 4096 channels and processing software. The measuring principle is based on the infinite energy extrapolation method, developed at the CEA 1. Reference 1 - J. Morel et al, Adaptation of the Gamma Spectrometry Method Based on the Infinite Energy Extrapolation to the Measurement of Small Amounts of Plutonium in Waste, ESARDA, Rome, Italy, The self-absorption phenomenon corresponds to the photon absorption by the material of its own rays. This is more pronounced when the atomic numbers of the constituent elements are high. One or two calculation stages will be required ( to estimate the spectrum. DIGITAL SIMULATIONS The MCNP code determines the energy spectrum of the photons detected in the germanium crystal (figure 2) of the detector fitted to the PLUM. This energy deposit of electrons set in motion by the detector photons is calculated, for one emitted photon, from: a 3D description of the detection system formed by the detector, its stand and collimator, in its measuring environment (figure 3). This modeling incorporates the physical and chemical (density and stoichiometry) and nuclear (cross sections) data characteristic of each material; a geometrical description of the photon source, its location in relation to the detector and the definition of its emission. Depending on how complex the photon source is (calibration source or nuclear material), one or two calculation stages will be required to estimate the spectrum. In the case of nuclear material, the effect of self-absorption 2 prompted us to carry out a two-stage calculation process for reasons of statistics and computing time. This process comprises calculating photon flow from 110 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

3 the radioactive object at the input side of the detector, then calculating the detector s response to this flow in normal incidence.

This calibration was carried out experimentally using a certified 152 Eu source, which was particularly useful because its γ emission spectrum is spread over an energy interval ranging from 121-1408

The model is improved by giving a very precise description of the detector s shape (cone-shaped, figure 2, not cylindrical as indicated on the drawings) and by adjusting the peripheral dead zone 2

3 3 the radioactive object at the input side of the detector, then calculating the detector s response to this flow in normal incidence. WHAT DIGITAL SIMULATIONS ACHIEVE The first modeling validation phase entails establishing the efficiency calibration curve 1 of the gamma-ray spectrometry system. This calibration was carried out experimentally using a certified 152 Eu source, which was particularly useful because its γ emission spectrum is spread over an energy interval ranging from kev (figure 4). The model is improved by giving a very precise description of the detector s shape (cone-shaped, figure 2, not cylindrical as indicated on the drawings) and by adjusting the peripheral dead zone 2 thickness in the manufacturer s recommended range to obtain constant deviation for all the 152 Eu lines. It produced the calibration curve presented in figure 5. As the relative deviations between the experimental efficiencies and those set by MCNP were between -2% and +4%, for the main 152 Eu lines, the detector model could be validated. Illustration of the spectra (figure 4) indicates that the total absorption peaks correlate well. In contrast, the Compton continuum level, due to the diffused rays, was underestimated because insufficient allowance was made for experimental background noise (such as naturally occurring potassium in the walls). In a second preparatory phase prior to waste drum modeling, the spectral response of the PLUM detector was simulated for a 1 g metal plutonium calibration source placed 1 m from the front panel of the detector. Given the significant self-absorption by plutonium, very accurate knowledge of its geometry would have been needed to model this phenomenon properly and thus validate the model. However the search for equivalent plutonium geometry (in line with the experiment) enabled us to study the scope of the phenomenon. Figure Eu spectrum. 1 - The efficiency of the detector, at a given energy and a given source-detector distance, is the ratio of the number of pulses recorded below the total absorption peak to the number of photons emitted by the source to this energy. 2 - The peripheral dead zone of the crystal, made up of germanium and diffused lithium, is one of the detector s electrodes. Security of radioactive materials and transport Figure 3 Figure 5 Axial section. Efficiency curve. SCIENTIFIC AND TECHNICAL REPORT

4 A third study involved a pot of PuO 2 powder with a mass of 500 g and well-characterized geometry, placed 1.5 m from the detector. The results obtained on the net surface of total absorption peaks of the plutonium, present relative simulation/experiment deviations that vary from +6% to +19% for energy levels in the range 129 to 450 kev. The simulation/experiment correlation is good, particularly as calculation in normal photon incidence in the crystal overestimates the results by 7% against the real geometry of the crystal at 1.5 m. Underestimation of the line at 59.6 kev of 241 Am may originate from residual deviation between the model and reality on the geometry of the pot (due to conflicting information on the drawing), the detector or the characteristics of the cadmium screen (measurements with a calibrating source emitting photons below 100 kev are required). The digital simulation results show that the calculated spectra are in line with the experimental spectra, in the case of a 152 Eu calibration source and a container of PuO 2. Subsequently, the PLUM device modeling will be applied to the calculation of spectra relating to real radioactive waste packages. The validation phase will involve increasingly complex measurement modeling to make allowance for the container, the matrix comprising the drum (physical and chemical composition, density, homogeneity), the radioactive material (activity level, position) and the presence of other more intense gamma-ray emitters (fission and activation products). Figure 6 CHACAL device. Digital simulation applied to the CHACAL passive neutron counter By digitally simulating neutron transfer in the CHACAL device using the MCNP code we can estimate the neutrons emitted by any neutron source placed inside the measuring cavity and detected in the device. The neutrons emitted following a fission reaction and detected in the device are calculated by digital simulation using the MCNP-PTA (Pulse Train Analysis) code, developed on the basis of the MCNP code at the European Commission Joint Research Center at Ispra (JRC/Ispra). The neutrons emitted following a fission reaction and detected in the device are calculated by digital simulation using the MCNP-PTA (Pulse Train Analysis) code. ( THE CHACAL DEVICE (CHAMBER FOR ELONGATED CONTAINERS) The CHACAL passive neutron-counting device (figure 6) is used to measure the plutonium held in large dimension containers (typical HxD = 75x16 cm), with a mass range from 50 g to 10 kg. It comprises a measuring chamber and an analyzer sensitive to coincident neutrons. The neutron measuring chamber is a thermal neutron well counter using twelve 3 He detectors in high-density polyethylene (HDPE). The sample cavity is shrouded with a cadmium liner designed to prevent re-entry of thermal neutrons. The passive measuring method is based on detecting the neutrons that accompany spontaneous fissions and fissions induced by 112 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

elements such as the oxygen of the PuO 2.

5 3 plutonium isotopes. These neutrons that are representative of the plutonium mass must be separated from those originating otherwise, mainly reactions (α, n) produced by plutonium isotope α-particle decay and on light elements such as the oxygen of the PuO 2. Using a method that analyzes temporal coincidences, a specific electronic device can separately quantify all the emitted and detected neutrons (known as singlets) and those that are representative of fissions, emitted by clusters and correlated over time, detected by pairs (known as doublets). The quantity of plutonium present in a container is deduced from the experimental neutron counts (singlets and doublets) using an analytical method developed by Hage and Cifarelli 1. Figure 7 Axial section. ( A specific electronic device can separately quantify all the neutrons emitted and detected and those that are representative of fissions. DIGITAL SIMULATIONS The MCNP computing code was used to establish the neutron detection efficiency in the CHACAL device s twelve 3 He detectors. It calculates the number of neutron captures (n, p) occurring in the active parts of the twelve detectors, for a neutron emitted by the neutron source from: a 3D description of the measuring chamber (figures 7 and 8), including the physical and chemical (density and stoichiometry) and nuclear (cross sections) characteristics of the materials; a geometrical description of the neutron source, its location in the sample cavity and the definition of its emission. It is easy to estimate the rate (in counts per second) obtained simulating singlets for this source or nuclear material placed in the CHACAL device as we know the intensity and energy distribution of neutron emission of the source or nuclear material in question. Reference 1 - D. M. Cifarelli and W. Hage, Models for a three Parameters Analysis of Neutron Signal Correlation measurements for Fissile Material Assay, Nuc. Instr. and Methods A251, 1986 Security of radioactive materials and transport Figure 8 Radial section. SCIENTIFIC AND TECHNICAL REPORT

Furthermore, simulation is made of the detection electronics relating to the counting chamber.

6 Figure 9 Neutron lifetime. The MCNP-PTA code operates on the basis of a conventional MCNP calculation in which fission reaction modeling is enhanced to make allowance for the full distribution of the neutron multiplicities. Furthermore, simulation is made of the detection electronics relating to the counting chamber. As we know the intensity and energy distribution of the neutron source, we can calculate the rate of singlets and doublets provided by the neutron counting chamber and analyzer detection set. Figure 10 Axial profile of singlets. Figure 11 Axial profile of doublets. WHAT DIGITAL SIMULATIONS ACHIEVE The digital simulation process goes through an initial phase to assess the quality of the model. This entails comparing the characteristic parameters measured from the CHACAL device against the simulated parameters: detection efficiency and lifetime of the neutrons in the system. The experimental device characterization measurements consisted in establishing these instrumental parameters using a calibration source of 252 Cf. Its emission energy spectrum is similar to that of plutonium isotopes. The neutron counts (singlets and doublets) are mapped in the sample cavity so that an application zone for the quantification method can be outlined. Figure 9 presents the evolution of doublets over time, obtained experimentally and by simulation with MCNP-PTA, for a source of 252 Cf centered in the sample cavity. At one moment t, the number of neutrons disappearing by unit of time is proportional to the number of neutrons present in the device. Thus the number of neutrons present in the device drops exponentially over time with a mean lifetime of λ. Figures 10 and 11 show the axial mapping of neutron counts (singlets and doublets), obtained experimentally and by MCNP- PTA simulation. In terms of absolute value, the simulations are consistent with the experiment. The singlet and doublet rate mapping demonstrates a constant efficiency zone over a height of 75 cm, thus enabling us to quantify the plutonium present in elongated containers. At the ends of the device, the count profile is explained by the presence of polyethylene reflectors. The effect on the doublets is lower because the neutron lifetime is longer than in the center. The counts can be standardized in the central zone because there is a sheet of cadmium (thermal neutron absorber) surrounding the moderator-counter sets. 114 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

7 3Security of radioactive materials and transport The uncertainty surrounding digital simulations is assessed by parametric sensitivity studies of the nuclear (spectra, cross sections), geometric, environmental data. ( and The uncertainty surrounding digital simulations is assessed by parametric sensitivity studies of the nuclear (spectra, cross sections), geometric, and environmental (repository chamber) data. A measurement campaign was carried out for the purpose in the PERLA (PERformance LAboratory) reference laboratory at JCR/Ispra, with certified neutron emission sources. The MCNP simulation/experiment deviations obtained on detection efficiency for the sources of 252 Cf, PuGa and PuO 2 (emission through spontaneous fission) and AmLi, AmB, AmF and AmBe ((α,n) emission) vary from -2% to +2%. The AmBe source was an exception as its observed deviation was +9%. In the last case, the geometric and physical and chemical data should be specified to improve the model of the source, (α,n) and (n,2n) reaction site. In a second research phase, calculation by simulating the neutron count rates may be performed on pots of nuclear material. Neutron counts (singlets and doublets) with samples of PuO 2 powder in a mass range of g, two samples of MOX powder and a reference BWR assembly have been acquired in the PERLA laboratory. The rate of singlets and also the sample s neutron multiplication factor were estimated by digital simulations built using the MCNP code. As a general rule, the digital simulations were in line with the experiment as the relative MCNP simulation/experiment deviations were less than 2% on the singlet rates and 5% on the neutron multiplication factors. The preliminary results obtained using the MCNP-PTA code (doublets) cover the 252 Cf calibration sources. It would be useful to continue the simulation process using samples of plutonium-bearing material. In order to do so, nuclear data relating to plutonium isotopes (specific neutron emissions, spectra of spontaneous and induced neutron emission, distributions of neutron multiplicities) must be incorporated into the MCNP-PTA code. These models will assist in analyzing the discrepancies between the operators statements and the measurements taken by the national inspection body, by enabling parametric sensitivity studies to be made of the singlets and doublets in relation to the characteristics of the nuclear material measured (total mass, density, physical and chemical form) and in the case of waste drums, in relation to the matrix characteristics. SCIENTIFIC AND TECHNICAL REPORT

8 1 Conclusions The results obtained using the MCNP and MCNP-PTA codes demonstrate the efficiency of digital simulation for estimating and confirming the results of measurements taken with neutron and gamma-ray measuring devices. Some experimental anomalies have been revealed (source calibration, instrumental parameters of the geometry, materials type). Conversely, the experimental measurements have uncovered nuclear and geometric data modeling errors. The deviations between the calculated values and the experimental results are as a rule below 10%. The models are being enhanced, using experimental measurements and parametric studies designed to pinpoint the calculation uncertainties linked to input parameters, with a view to using them on nuclear material samples controlled during inspections. Acknowledgements We should like to thank Bertrand Pérot (CEA), Paolo Peerani and Marc Looman (JCR/Ispra), and Hervé Vidal (trainee ULP-Strasbourg), for their active participation in these studies. ( The experimental measurements have uncovered nuclear and geometric data modeling errors. 116 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE

Rρ corresponds to the doublets counting

Rρ corresponds to the doublets counting Some techniques applied for plutonium measurements in waste drums B. Autrusson, J.L. Dufour, P. Funk, T. Lambert, N. Pépin, B. Thaurel Institut de Protection et de Sûreté Nucléaire, BP 6, 92265 Fontenay-aux-Roses,