The Pennsylvania State University. The Graduate School. College of Engineering SUPPORTING ANALYSIS FOR THE DEVELOPMENT OF A RECONFIGURABLE

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1 The Pennsylvania State University The Graduate School College of Engineering SUPPORTING ANALYSIS FOR THE DEVELOPMENT OF A RECONFIGURABLE DIPOLE CHAFF ELEMENT FOR THE REMOTE PASSIVE DETECTION OF NEUTRONS A Thesis in Nuclear Engineering by Brian A. Vresko 2012 Brian A. Vresko Submitted in Partial Fulfillment of the Requirements for the Degree of Master of Science August 2012

2 ii The thesis of Brian A. Vresko was reviewed and approved* by the following: Jack S. Brenizer J. Lee Everett Professor of Mechanical and Nuclear Engineering Thesis advisor David C. Swanson Senior Research Associate at Penn State University Arthur T. Motta Professor of Nuclear Engineering and Materials Science and Engineering Chair of Nuclear Engineering *Signatures are on file in the Graduate School.

3 iii Abstract The objective of thesis research was to evaluate potential candidate materials that were under consideration in the development of a new neutron detecting device. This project was inspired by a detector that consisted of reconfigurable dipole chaff elements for the remote passive detection of chemical agents. It was proposed that a similar detector could be developed to detect neutrons instead of chemicals, by modifying radiosensitive material and adding a neutron converter. The radiosensitive material that was being considered for the neutron detector design was the diacetylene, PCDA. The three different converter materials that were considered for the detector design were gadolinium metal, boron nitride, and lithium silicate. A number of important observations were made from the experiments performed in this study. The results obtained from the gamma irradiation experiments indicated that the incorporation of gadolinium in the PCDA samples increased the materials gamma-ray sensitivity. This increased sensitivity was not desirable, since a goal for the detector design was for it to be insensitive to background radiation. The neutron irradiation experiments showed that the radiosensitive material should not be thicker than the range of the ions or charged particles, generated by a neutron-converter interaction. Additionally, there should be a distinct layer of converter material in the detector design, instead of a distribution of converter nanoparticles in the radiosensitive material. Because the amount of neutron converter that can be incorporated in the PCDA is limited, utilizing a discrete layer of converter increases the number of converter atoms. For a given flux, this yields an increased number of neutron absorptions, and therefore increases the energy that can be emitted from the converter into the PCDA. Models of the converter and the radiosensitive material were utilized to determine the neutron interaction distributions in the converter and the reaction product interaction in the radiosensitive

4 iv material. The Monte Carlo methods based computer programs CASINO and SRIM were used to model the produced electrons and heavy charged particles, respectively. Because the converters were thin, neutron interactions within the converters were modeled assuming simple exponential attenuation. The modeling indicated that different converter materials should be incorporated into the detector design depending on the design limitations and how the device is to be deployed. If boron nitride is used as the converter material in the detector design, the PCDA radiosensitive material thickness should be set to 5 microns and the boron nitride thickness should be equal to or greater than 10 microns. When gadolinium metal foil is used as the converter, the PCDA radiosensitive material should be set to 50 microns and the gadolinium thickness should be equal to or greater than 10 microns. In the case where lithium silicate is used as the converter material, the radiosensitive material thickness should be set to 50 microns and the converter material thickness should be set to 50 microns or greater. For all scenarios mentioned above, the converter material should be placed behind the radiosensitive material. This allows the thickness of the converter material to be increased without any penalty to the energy emitted from the converter material, as long as the alignment of the device is maintained. Enriching the boron and the lithium in their neutron absorbing isotopes is also recommended, since using enriched converters greatly increases the energy emitted from the converter material. It is important to note that the recommended converter thicknesses above would need to be modified if the radiosensitive material in the final detector design does not have the optimum density of 0.92 g/cm 3. Based on the experimental and modeling results, enriched boron nitride or enriched lithium silicate should be used as the converter materials. If the detector design does not allow for the recommended thickness mentioned above, then a different converter material should be used based on the design limitations.

5 v Table of Contents List of Tables... vii List of Figures... viii Acknowledgements... xii Chapter 1. Introduction... 1 Chapter 2. Background and Theory Neutron Source Materials Radiosensitive Materials Converter Materials Detector Designs Chapter 3. Experimental Testing Data and Results Gold Foil Experiment Cobalt-60 Pool Irradiation Facility Experiments Neutron Beam Experiments Chapter 4. Modeling of Converters and Radiosensitive Materials Neutron Absorption in Gadolinium Modeling Methodology Gadolinium Metal Analysis with Average Energy Electrons Average Energy Electrons versus Multiple Energy Electrons Neutron-Gadolinium Interaction Products in PCDA PCDA Density Analysis with Electrons Boron Nitride and Lithium Silicate Analysis Overview Boron Nitride Analysis Neutron-Boron Interaction Products in PCDA Lithium Silicate Analysis Neutron-Lithium Interaction Products in PCDA... 57

6 vi PCDA Density Analysis with Heavy Charged Particles Comparison of Converters with Natural Isotopic Abundance Comparison of Converters with Enriched Isotopic Abundance Chapter 5. Conclusions and Future Work Summary and Conclusions Future Work References... 85

7 vii List of Tables Table 2-1: IAEA Significant Quantities for a number of nuclear materials Table 2-2: Neutron production, by means of spontaneous fission, for common nuclides Table 2-3: Neutron production from significant quantities of uranium, at various enrichments Table 2-4: The natural abundance and thermal neutron capture cross-section of all stable gadolinium isotopes Table 2-5: Energy and frequency of emission for internal conversion electrons released from neutrongadolinium interactions. These internal conversion electrons are the primary contributors to neutron radiography Table 3-1: The exposures performed on each sample during the second gamma irradiation experiment. Each exposure is listed with the approximate dose that the samples received, and the total accumulated dose Table 3-2: The exposures performed on each sample during the first neutron irradiation experiment. Each exposure time is listed, along with the associated neutron dose and gamma dose Table 4-1: Energy and frequency of emission for internal conversion electrons released from neutron gadolinium interactions. These internal conversion electrons are the primary contributors to neutron radiography. Entries that were used in the analysis are highlighted in yellow Table 5-1: Calculation of average energy per interaction deposited in the radiosensitive material, for each converter type placed behind the radiosensitive material. Also listed in the table is the penetration depth and width of the energy distribution in PCDA for each converter type

8 viii List of Figures Figure 2-1: Illustration of topochemical polymerization of a diactylene Figure 2-2: Illustration of two initial material designs considered by the group. The illustration is shown with the radiosensitive material PTS and the converter material gadolinium Figure 3-1: Neutron-gamma interaction cross section of gold versus the incident neutron energy Figure 3-2: Flux of neutrons, with energies between 0 ev and 1 kev, detected on a 0.25 g pure gold foil versus the thickness of polyethylene placed between a 2500kg mass of natural uranium and the foil Figure 3-3: The sample holder used for the gamma irradiation experiments. Samples were taped to the outer surface between 2.54 cm and cm from the bottom of the PVC pipe Figure 3-4: Comparison of the PCDA/PMMA samples containing boron and gadolinium nanoparticles after gamma irradiation. The red (boron) and blue (gadolinium) dashed circles indicate similar levels of polymerization Figure 3-5: Comparison of the PCDA/PVA samples containing boron and gadolinium nanoparticles after gamma irradiation. The red (boron) and blue (gadolinium) dashed circles indicate similar levels of polymerization Figure 3-6: Photograph of the sample holder used in the neutron irradiation experiments. Samples were mounted on paper or silicon wafers then taped to a piece of cardboard with kapton tape Figure 3-7: Comparison of the PCDA/PMMA samples containing gadolinium and boron nanoparticles, irradiated during the first neutron experiment Figure 3-8: Illustration of how sample thickness affects polymerization. The blue circles represent the area of energy deposition by the emitted particles. This effect was observed in the samples irradiated during the first neutron experiment Figure 3-9: A Comparison of PCDA/PMMA samples containing gadolinium and boron nanoparticles, irradiated during the second neutron experiment Figure 4-1: The fraction of neutrons absorbed in natural gadolinium versus the gadolinium thickness. Equation 4-2 was used to calculate the fraction of neutrons absorbed in natural gadolinium Figure 4-2: Illustration of modeling methodology used in the gadolinium metal analysis. The converter material was broken into 1 micron layers, and it was assumed that all neutron-converter interactions occurred in the center of each layer

9 ix Figure 4-3: Illustration of the first detector configuration with the converter material placed between the radiosensitive material and the neutron source Figure 4-4: Illustration of the second detector configuration with the radiosensitive material between the converter material and the neutron source Figure 4-5: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the first configuration versus the gadolinium metal foil thickness. The electrons are emitted from the gadolinium converter with an average energy of 72 kev Figure 4-6: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the second configuration versus the gadolinium metal foil thickness. The electrons are emitted from the gadolinium converter with an average energy of 72 kev Figure 4-7: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the first configuration versus the gadolinium metal foil thickness. The average energy data is represented by blue diamonds and the multiple energy data is represented by red squares Figure 4-8: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the second configuration versus the gadolinium metal foil thickness. The average energy data is represented by blue diamonds and the multiple energy data is represented by red squares Figure 4-9: The electron track plot generated in CASINO, for 72 kev electrons in PCDA Figure 4-10: The electron distribution plot generated in CASINO, for 72 kev electrons in PCDA Figure 4-11: Maximum electron energy penetration depth of 72 kev electrons in PCDA versus the density of the PCDA radiosensitive material Figure 4-12: The fraction of energy emitted from the boron nitride converter layer of a detector oriented in the first configuration (Figure 4-3) versus the boron nitride thickness. The helium ion is emitted with an energy of 1.47 MeV and the lithium ion is emitted with an energy of 0.84 MeV Figure 4-13: The fraction of energy emitted from the boron nitride converter layer of a detector oriented in the second configuration (Figure 4-4) versus the boron nitride thickness. The helium ion is emitted with an energy of 1.47 MeV and the lithium ion is emitted with an energy of 0.84 MeV Figure 4-14: Ion track plots generated in SRIM, for 1.47 MeV helium ions in PCDA with a density of 0.92 g/cm Figure 4-15: Ion track plots generated in SRIM, for 0.84 MeV lithium ions in PCDA with a density of 0.92 g/cm

10 x Figure 4-16: The fraction of energy emitted from the lithium silicate converter layer of a detector oriented in the first configuration versus the lithium silicate thickness. The hydrogen ion is emitted with an energy of 2.75 MeV and the helium ion is emitted with an energy of 2.05 MeV Figure 4-17: The fraction of energy emitted from the lithium silicate converter layer of a detector oriented in the second configuration versus the lithium silicate thickness. The hydrogen ion is emitted with an energy of 2.75 MeV and the helium ion is emitted with an energy of 2.05 MeV Figure 4-18: Ion track plots generated in SRIM, for 2.75 MeV hydrogen ions in PCDA with a density of 0.92 g/cm Figure 4-19: Ion track plots generated in SRIM, for 2.05 MeV helium ions in PCDA with a density of 0.92 g/cm Figure 4-20: Maximum ion penetration depth of boron-neutron interaction products in PCDA versus the density of the PCDA radiosensitive material Figure 4-21: Maximum ion penetration depth of lithium-neutron interaction products in PCDA versus the density of the PCDA radiosensitive material Figure 4-22: The fraction of energy emitted from the converter layer of a detector oriented in the first configuration versus the converter thickness. The data from the gadolinium metal, boron nitride, and lithium silicate converters are shown on the figure Figure 4-23: The fraction of energy emitted from the converter layer of a detector oriented in the second configuration versus the converter thickness. The data from the gadolinium metal, boron nitride, and lithium silicate converters are shown on the figure Figure 4-24: The average energy emitted from the converter layer of a detector oriented in the first configuration versus the converter thickness. The data from the gadolinium metal, boron nitride, and lithium silicate converters are shown on the figure Figure 4-25: The average energy emitted from the converter layer of a detector oriented in the second configuration versus the converter thickness. The data from the gadolinium metal, boron nitride, and lithium silicate converters are shown on the figure Figure 4-26: The fraction of energy emitted from the converter layer of a detector oriented in the first configuration versus the converter thickness. The data from the gadolinium metal, boron nitride (enriched to 100% boron-10), and lithium silicate (enriched to 100% lithium-6) are shown on the figure

11 xi Figure 4-27: The fraction of energy emitted from the converter layer of a detector oriented in the second configuration versus the converter thickness. The data from the gadolinium metal, boron nitride (enriched to 100% boron-10), and lithium silicate (enriched to 100% lithium- 6) are shown on the figure Figure 4-28: The average energy emitted from the converter layer of a detector oriented in the first configuration versus the converter thickness. The data from the gadolinium metal, boron nitride (enriched to 100% boron-10), and lithium silicate (enriched to 100% lithium-6) are shown on the figure Figure 4-29: The average energy emitted from the converter layer of a detector oriented in the second configuration versus the converter thickness. The data from the gadolinium metal, boron nitride (enriched to 100% boron-10), and lithium silicate (enriched to 100% lithium-6) are shown on the figure

12 xii Acknowledgement I would like to thank the Defense Threat Reduction Agency for its support of this research, which was funded by DTRA01-03-D Neutron Detection under the direction of Dr. Timothy Leong.

13 1 Chapter 1 Introduction This study was part of a joint project of members of Penn State s Electrical Engineering Department, Applied Research Laboratory (ARL), and Nuclear Engineering Program, with the goal of developing a new neutron detecting device. The final design of the detector was proposed to be a small, affordable device that could detect neutron fluxes greater than background levels. The idea for the device came from a chemical detector that was developed at The Pennsylvania State University. These detectors were reconfigurable dipole chaff elements for the remote passive detection of chemical agents. In this design, one dipole is used as a reference, and the other dipole is used for detection of the chemical agent. [1] As described in Reference 1, the detection dipole is the same size as the reference dipole with a chemoresistive switch incorporated into it. Before exposure, the detection dipole has the same radar cross section as the reference dipole. When exposed to the proper chemical agent, the switch changes from a state of low conductivity to a state of high conductivity. After exposure, the change in conductivity makes the detection dipole electrically longer and causes a shift down in frequency in the radar cross section. Rather than detecting a chemical agent, like the device described in Reference 1, this project attempted to design a dipole chaff element for the passive detection of neutrons using radar. The focus of the work presented in this thesis was to guide the selection of a converter material to be used in the detector design through modeling and experimental testing. Additionally, these analyses provided data to select the proper dimensions and orientation of various components used in the detector device. Detecting neutrons is a particularly difficult problem. Since neutrons are neutral particles they do not interact readily with a variety of materials. Also, the proposed radiosensitive materials are

14 2 intended to be thin, on the order of microns thick. This causes an additional complication that reduces the radiosensitive material s ability to interact with radiation. The interaction between uncharged radiation (neutrons and gamma rays) and a particular material is probabilistic, so as the material thickness increases there is a higher probability of interaction. To increase the sensitivity to neutrons, a converter material was incorporated into the detector design. Common converter materials that were considered for use in the detector were gadolinium, boron and lithium. These converter materials have isotopes that have both a high neutron absorption cross section, and release charged particles when they interact with neutrons that can interact readily with the radiosensitive material. These converter materials, and their potential incorporation into the detector designs, are discussed in depth in Chapters 2. Analysis of these converter materials through experimentation and computer modeling are discussed in Chapters 3 and 4. The facilities at The Pennsylvania State University were used to test various components of the detector, and detector prototypes. Gamma irradiation facilities were used to experimentally determine how much radiation was needed to polymerize the radiosensitive material. These experiments were important because it was desirable for the final detector device to be insensitive to gamma radiation, which would reduce the probability of having background radiation cause a false positive. The Breazeale Nuclear Reactor was used to test the sensitivity of detector prototypes to thermal neutron radiation.

15 3 Chapter 2 Background and Theory The following section discusses background and theory used in support of the development of the neutron detection device. This section reviews the neutron sources that were considered for testing and the potential sources that were available for the research. Additionally, this section covers the various materials that were considered for use in the detector, and how those materials would be arranged in the detector Neutron Source Materials The main goal of this research project was to design a detector that can detect neutrons emitted, by spontaneous fission, from a quantity of nuclear material. Once the design was finalized, the International Atomic Energy Agency (IAEA) list of significant quantities was to be used as a basis to determine appropriate neutron flux levels for testing the detector. A significant quantity of nuclear material is defined as the approximate amount of nuclear material for which the possibility of manufacturing a nuclear explosive device cannot be excluded. [2] Significant quantities take into account unavoidable losses due to conversion and manufacturing processes and should not be confused with critical masses. Table 2-1 lists a number of nuclear materials and their corresponding IAEA significant quantity.

16 4 Table 2-1 IAEA Significant Quantities for a number of nuclear materials. [2] Nuclear Material Significant Quantity (kg) Pu 8 U HEU (U %) 25 U (U-235 < 20%) 75 (or 10,000 natural U) (or 20,000 depleted U) Th 20,000 This research focused on uranium, with various levels of enrichment, for testing and modeling. Uranium was chosen primarily due to the availability of source materials at The Pennsylvania State University. The facilities at the Radiation Science and Engineering Center and the Breazeale Nuclear Reactor had a number of low enriched uranium sources that were available for use in the tests. The Breazeale Nuclear Reactor is a 1MW TRIGA, pool-type reactor. The reactor operates using TRIGA assemblies, made of uranium zirconium hydride fuel, enriched to 8.5% or 12% U-235. Fuel assemblies that have not yet been used in the reactor are stored at the facility, and were available as a potential neutron source at the time of the research. Another source that could be used for testing was a storage assembly of natural uranium fuel rods, at the Radiation Science and Engineering Center. The assembly consists of a wooden rack and a number of rods containing approximately 2500 kg of natural uranium, arranged with a triangular pitch. The natural uranium fuel rods are encapsulated in thin walled aluminum tubing, with an outside diameter of approximately 3.3 cm. Potential detector materials were also able to be tested in the neutron beam facility at the Radiation Science and Engineering Center. During operation of the Breazeale Nuclear Reactor, some of the neutrons emitted from the core enter a beam collimator in Beam Port 4, creating a well-collimated thermal neutron beam. The intensity of the neutron beam can be adjusted using reactor power, and the

17 5 beam can be used to imitate the approximate neutron flux that is emitted from a significant quantity of low enriched uranium. Uranium can decay by either spontaneous fission or by alpha decay. The alpha particles are difficult to detect. This is because the heavy charged particles interact readily with matter. Most of the alpha particles are absorbed before they escaped the special nuclear material itself and those that do escape would be absorbed in a few centimeters of air. When uranium undergoes spontaneous fission it emits a small number of neutrons with a spectrum of energy. Theses neutrons can potentially be detected at a distance since they have a much lower probability of interaction with matter than the alpha particles. Each isotope of special nuclear material has a different spontaneous fission rate. Table 2-2 shows the spontaneous neutron production rates of typical isotopes of special nuclear material. Table 2-2 Neutron production, by means of spontaneous fission, for common nuclides. [3] Neutron Production by Spontaneous Fission Nuclide Fission Half-Life α-decay Half-Life Neutron Production (years) (years) (neutrons/sec/gram) U * *10 8 yrs 8.0*10-4 U * *10 9 yrs 1.6*10-2 Pu * *10 4 yrs 3.0*10-2 Pu * *10 3 yrs 1.0*10 3 Cf * *10 0 yrs 2.3*10 12 The data from Table 2-1 and 2-2 was used to calculate the approximate neutron production of uranium with various enrichments of uranium-235 (U-235). These neutron production numbers were calculated using the IAEA significant quantities for selected enrichments. The neutron flux was also calculated for each enrichment at a distance of one meter from the source. Table 2-3 shows the neutron production and the neutron flux, at one meter from the source, for the selected U-235 enrichments.

18 6 Table 2-3 Neutron production from significant quantities of uranium, at various enrichments. Nuclear Material Mass Neutron Production Neutron Flux at 1 m (kg) (n/s) (n/cm 2 /s) U (0.7% U-235) 9.07E E E+00 U (5% U-235) 7.50E E E-03 U (10% U-235) 7.50E E E-03 U (20% U-235) 2.50E E E-03 U (40% U-235) 2.50E E E-03 U (60% U-235) 2.50E E E-03 U (80% U-235) 2.50E E E-04 U (100% U-235) 2.50E E E-04 It can be observed from the above table that as the enrichment of U-235 increases in the uranium source, the neutron production and neutron flux decreases. This is due to the different spontaneous fission rates of the uranium isotopes. The spontaneous fission rate of U-235 is two orders of magnitude lower than that of U-238, and as the concentration of U-235 increases the neutron production rate decreases. In practical terms this indicates that as the enrichment of U-235 increases, it becomes increasingly difficult to detect the uranium source by neutron detection Radiosensitive Materials For the proposed detector to change state in the presence of neutrons, a radiosensitive material had to be incorporated into the design. Two radiosensitive materials were tested in support of this research, bis-(p-toluene sulfonate) (PTS) and Pentacosa-10,12-diynoic acid (PCDA). Radiosensitive materials start out as monomers, and when energy is imparted into the monomers they link together to form polymers chains. This process of cross-linking monomers into polymers is called topochemical

19 7 polymerization of a diacetylene. Figure 2-1 illustrates the process of changing the radiosensitive material from a monomer to a polymer, also known as the topochemical polymerization of a diacetylene. Figure 2-1: Illustration of topochemical polymerization of a diactylene. [4] When the radiosensitive materials change from a monomer to a polymer, some of their properties change. The property of interest for this research was the material s permittivity. Permittivity is the ability of a material to store electrical potential energy under the influence of an electric field measured by the ratio of the capacitance of a capacitor with the material as dielectric to its capacitance with vacuum as dielectric. [5] Since permittivity is related to capacitance, this material property is important for antenna designs. When two antenna arrays are connected by a material, the permittivity of the connecting material determines how the antennas reflect radar. Depending on the permittivity, the antennas will reflect radar as two discrete antennas or they will be connected electrically and reflect radar as if they are one large antenna. If a radiosensitive material is incorporated into an antenna design, the change in permittivity during polymerization can be utilized to change how an antenna device reflects radar.

20 8 Colleagues working to establish the electrical properties of the sensitive materials found that the commercially available radiosensitive materials were difficult to deposit onto a smooth dry surface. In an effort to get the material deposited onto a surface for testing, they mixed the radiosensitive materials with a number of commercially available binders including paraffin wax, PVA, and PMMA. These samples were exposed to gamma and neutron radiation and the results of those experiments are discussed in Chapter Converter Materials Since the proposed radiosensitive materials do not readily absorb neutrons, another material needed to be incorporated into the detector design to increase sensitivity. Converter materials were chosen because they easily interact with neutrons and emit charged particles. These charged particles can then exit the converter material and interact with the radiosensitive material. Gadolinium, boron, and lithium are commonly used as neutron converter materials that were considered for the detector design. In this study, all neutrons that are incident to the converter material were assumed to be thermal and have an energy of ev. The converter materials considered for the detector, interact more readily with thermal neutrons than with higher energy neutrons. When detector materials are tested in a thermal neutron flux, desired responses will occur with shorter irradiation times. Additionally, assuming that all neutrons are emitted at thermal energies, instead of a spectrum of energies, simplified the modeling analyses performed in Chapter 4.

21 9 Gadolinium has a number of stable isotopes that interact with thermal neutrons. Table 2-4 shows the natural abundance and the thermal neutron ( ev) capture cross section for all stable isotopes of gadolinium. Table 2-4 The natural abundance and thermal neutron capture cross-section of all stable gadolinium isotopes. [6] Gadolinium Isotope (mass number) Natural Abundance (%) Thermal Neutron Capture Cross Section (barns) E E E E E E E+0 When a neutron interacts via a capture reaction with gadolinium, it can release a gamma ray or an internal conversion electron at one of thousands of different energies. [7] If an internal conversion electron is emitted, it can escape the converter material and deposit its energy into the radiosensitive material. Table 2-5 shows the energy and frequency of the internal conversion electrons that are the main contributors to exposing film in neutron radiography. Since neutron radiography utilizes energy deposition in a thin photographic film emulsion for creation of the radiographic image, it is reasonable to utilize these same energy conversion electrons. They will be the primary contributors to the energy deposition in the radiosensitive material in the proposed detector.

22 10 Table 2-5 Energy and frequency of emission for internal conversion electrons released from neutron-gadolinium interactions. These internal conversion electrons are the primary contributors to neutron radiography. [7] Energy Group Energy Frequency Range in Gd [CSDA] Energy/Interaction % of Energy (kev) (%) (μm) (kev) (%) Total: The isotope of boron that interacts with thermal neutrons is boron-10, and it has an absorption cross section of barns for ev thermal neutrons. [8] Natural boron is composed of boron- 10, with an abundance of 19.9%, and boron-11 with an abundance of 80.1%. [6] When a neutron interacts with a boron-10 atom, the resultant products can be emitted at one of two different energies. [9] Ninety four percent (94%) of the time the lithium ion is left in an excited state and the resultant products, the alpha particle and the lithium ion, leave with a total kinetic energy of MeV. The other 6% of the time the lithium ion is in its ground state and the resultant products leave with a total kinetic energy of MeV. For all of the boron analysis that is reported in the following sections, it was assumed that the products generated from a neutron-boron interaction were released with a total kinetic energy of MeV. The two possible neutron-boron-10 interactions are illustrated in Equations 2-1 and 2-2.

23 [94%] [6%] 2-2 The third converter material that was analyzed in this study was lithium. The isotope of lithium that interacts with thermal neutrons is lithium-6. It has an absorption cross section of barns for ev thermal neutrons. [8] Natural lithium is composed of lithium-6, with an abundance of 7.59%, and lithium-7 with an abundance of 92.41%. [6] When a neutron interacts with a lithium-6 atom, the resultant products (an alpha particle and a triton) are emitted with a total kinetic energy of 4.8 MeV. [10] The neutron-lithium-6 interaction is illustrated in Equation Detector Designs There were two configurations that were considered for the radiosensitive and converter material layers in the detector. The first configuration, referred to as Type I in this thesis, incorporated nanoparticles of a converter material into the radiosensitive material. The second design, referred to as Type II in this thesis, consisted of alternating layers of radiosensitive material and converter material. In both types, a fraction of the incoming neutrons interact with the converter material and release energetic charged particles that could then deposit their energy into the radiosensitive material. Figure 2-2 illustrates both of the initial material designs that were considered by the group.

24 12 Figure 2-2: Illustration of two initial material designs considered by the group. The illustration is shown with the radiosensitive material PTS and the converter material gadolinium. [4] Approximately 25 different combinations of samples, containing a mix of radiosensitive materials, binding materials, and converter materials, were tested in the facilities at the Research Science and Engineering Center. The experiments are discussed in more detail in Chapter 3.

25 13 Chapter 3 Experimental Testing Data and Results The following sections discuss the various tests that were performed in support of the development of the detector. Two gamma irradiation experiments were performed to determine the sensitivity of potential detector materials in a gamma flux. Additionally, two neutron irradiation experiments that were performed to determine the sensitivity of potential detector material in a neutron flux. The facilities at the Penn State Radiation Science and Engineering Center were used for both the gamma and neutron irradiation experiments. Additional experiments were performed at the Radiation Science and Engineering Center to determine the intensity of the neutron flux off of different uranium sources. This data was used to benchmark the MCNP neutron flux predictions. 3.1 Gold Foil Experiments A gold foil was utilized to measure the neutron flux levels emitted from the 2500 kg mass of natural uranium located in the Radiation Science and Engineering Center. Gold was used as a detecting material because it has a high cross section over a wide spectrum of neutron energies. It can be observed in Figure 3-1 that gold maintains a high neutron absorption cross section up to approximately 1 kev, where it decreases below 1 barn. [8]

26 14 Figure 3-1: Neutron-gamma interaction cross section of gold versus the incident neutron energy. When neutrons interact with gold-197, there is a chance that the neutron will be absorbed and create gold-198. Gold-198 has a half-life of 2.27 days and decays to mercury-198 by beta and gamma emission. [6] Over a few days of exposure on the mass of natural uranium, the gold became activated and ultimately reached a steady state in its activity. The gold foil was then removed from the flux, and the amount of activation was measured. A neutron flux was then determined depending on the activity of the gold foil. For the first gold foil irradiation, a 0.25 gram pure gold foil was placed on the mass of uranium. After eight days of irradiation, the activity on the gold foil was measured but no activation was detected. This is due to the fact that the neutrons coming off of the natural uranium assembly are primarily fast neutrons, and there was not enough low Z number material present between the mass of uranium and the detector to moderate the neutrons down to thermal energies.

27 15 To activate the gold, the thermal neutron flux was increases by placing a moderator between the mass of uranium and the foil. Polyethylene was chosen as the desired moderating material due to its high concentration of hydrogen. Analysis was performed to determine how thick the polyethylene needed to be to produce the most thermal neutrons. There are two competing factors present when using polyethylene as a moderator. As the thickness of the polyethylene block increases, more neutrons interact with the material and slow down to thermal energies. However, when the neutrons are moderated, they may scattered away from the target, or once the neutrons reach thermal levels they may be absorbed in the polyethylene material. MCNP was used to calculate the total neutron flux emitted from the mass of uranium, and the thermal neutron flux emitted from the mass of uranium when various thicknesses of polyethylene were place on it. Figure 3-2 shows the neutron flux, predicted by MCNP, for neutrons with energies between 0 ev and 1 kev. Figure 3-2: Flux of neutrons, with energies between 0 ev and 1 kev, detected on a 0.25 g pure gold foil versus the thickness of polyethylene placed between a 2500kg mass of natural uranium and the foil.

28 16 From Figure 3-2, the largest flux that could be detected by the gold foil measurement occurs when 4-5 cm of polyethylene is used as a moderator. With this knowledge, a second experiment was performed on the uranium source using the same type of gold foil, but this time a 5.08 cm polyethylene block was placed between the gold foil and the 2500kg mass of natural uranium fuel. The foil was again exposed for eight days. After the irradiation, the foil was removed and the activity on the gold foil was measured. This time activity was detected on the gold foil, and the total neutron flux coming off of the natural uranium assembly was determined to be 2.3 neutrons/cm 2 /s This is similar to the total neutron flux of 1.2 neutrons/cm 2 /s predicted by the MCNP modeling. For the MCNP flux predictions, the statistical error was less than 5% from 0cm to 15cm from the source, and less than 10% for 20cm and 25 cm from the source. It was expected that the mass of natural uranium would emit primarily fast neutrons, and the results of the gold foil experiment confirmed that assumption. Most materials are significantly more sensitive to thermal neutrons than fast neutron. This is a slight concern for the development of the proposed neutron detector. However, neutrons can be naturally moderated down to thermal levels as they travel through air. The distance that a neutron has to travel in air before being moderated to thermal energies is a relatively long distance, on the order of meters, and is dependent on a number of factors including the distance above sea level and the humidity. For the remaining neutron experiments performed on the proposed detector materials, the neutrons were moderated down to thermal levels. 3.2 Cobalt-60 Pool Irradiation Facility Experiments The initial experiments that were performed on the radiosensitive materials tested their sensitivity to gamma radiation. Insensitivity to gamma rays is important for a passive neutron detector

29 17 to help prevent the detector from giving a positive response strictly due to the background gamma radiation. The samples were irradiated with gamma rays using a cobalt-60 source. The cobalt bay at the Radiation Science and Engineering Center contains two dry irradiation tubes. The bottom of each tube is surrounded with cobalt-60 pencils. All experiments were performed using the large irradiation tube. The samples were mounted on a holder and lowered down the irradiation tube to the position of the cobalt-60 pencils. At various irradiation doses (controlled by the exposure time), the samples were removed and photographed to observe a color change in the radiosensitive material. This color change was an indication that the radiosensitive material had become polymerized and that the material properties of the sample had changed. It should be noted that the samples did not become radioactive during the gamma irradiation, so once the samples were removed from the irradiation tube they could be handled and transported to other laboratories for analysis. The first two gamma irradiation experiments tested approximately 20 samples in the pool cobalt irradiation facility. The samples contained the radiosensitive material PCDA, along with a binding material of paraffin wax, PMMA, or PVA. Additionally, some samples contained nanoparticles of the converter materials gadolinium or boron (replicating a Type I configuration). There were also a few samples in each experiment that had the radiosensitive material deposited on gadolinium metal foil (replicating a Type II configuration). The samples were mounted on a piece of PVC pipe and lowered to the bottom of the irradiation tube. Figure 3-3 shows a picture of the sample holder that was used for the gamma irradiations. Lines on the PVC pipe indicate the region on the sample holder that is between 2.54 cm and cm from the bottom of the cobalt irradiation tube. The samples were placed between 2.54 cm and cm from the bottom of the tube, because the gamma ray flux was very

30 18 uniform (+/-7% of the average) in that region. All samples placed in that region received approximately the same dose. Figure 3-3: The sample holder used for the gamma irradiation experiments. Samples were taped to the outer surface between 2.54 cm and cm from the bottom of the PVC pipe. For the first experiment, the average dose rate from the cobalt-60 gamma source was approximately 236 rad/min. All material samples received three doses of approximately 1 krad, then a dose of approximately 2 krad, then approximately 4 krad, and finally a dose of approximately 11 krad. The samples received a total dose of approximately 20 krad. In this experiment, the samples contained the radiosensitive material PCDA, along with a binding material of paraffin wax, PMMA, or PVA. Approximately 5 samples also contained nanoparticles of gadolinium. After each irradiation dose, the sample holder was removed from the tube and photographs were taken of each sample. Due to the lighting in the gamma ray irradiation facility, clear pictures were not obtained during the experiment.

31 19 However, the results of each irradiation were documented every time the samples were removed from the tube. Preliminary results indicated that some samples polymerized more quickly than others in the gamma ray flux. From this experiment it was observed that the samples containing paraffin wax and the samples containing gadolinium polymerized faster than samples that did not containing these components. With these results, it was important to determine if a different converter would cause the radiosensitive material to be less sensitive to gamma radiation. For the next gamma ray experiment, it was decided that additional samples would be made with boron nanoparticles. To get a clearer picture of how the samples changed with respect to each other, the irradiation time of the samples was decreased in order to decrease the magnitude of each dose step. The second gamma irradiation test was performed with cobalt-60 gamma rays using a variety of radiosensitive material, binder, and converter combinations. A total of 20 samples were simultaneously irradiated in this test. PCDA was again used as the radiosensitive material, but was combined with only two binding materials PVA or PMMA. Control samples were created that contained no converter material for comparison with samples that contained gadolinium nanoparticles, and samples that contained boron nanoparticles. For this experiment the average dose rate was measured to be approximately 236 rad/min between 2.54 cm and cm from the bottom of the cobalt irradiation tube. The same sample holder was used to mount the samples. During this experiment, smaller dose intervals were used to observe how the samples changed at lower gamma ray doses. It was important to refine the gamma dose sensitivity, because these same samples would eventually be tested in the neutron beam where a relatively small (compared to that in the cobalt-60 irradiation tube) gamma flux is present. If the samples polymerized with low gamma ray doses, it would not be known if the samples had polymerized due to the neutron flux or if they

32 20 polymerized due to the gamma ray flux, when they were being irradiated by the neutron beam. Table 3-1 shows the gamma ray exposures that each sample received. Table 3-1 The exposures performed on each sample during the second gamma irradiation experiment. Each exposure is listed with the approximate dose that the samples received, and the total accumulated dose. Exposure Exposure Dose Total Dose Number (rad) (rad) Figures 3-4 and 3-5 show the photographs of the samples taken after various gamma ray doses. The degrees of polymerization on the PCDA/PMMA samples containing boron and gadolinium nanoparticles were compared at various gamma ray doses. Similarly, the degrees of polymerization on the PCDA/PVA samples containing boron and gadolinium nanoparticles were also compared at various gamma ray doses. The dashed red and blue circles indicate samples that were determined, by the materials development team, to have similar levels of polymerization.

33 21 Figure 3-4: Comparison of the PCDA/PMMA samples containing boron and gadolinium nanoparticles after gamma irradiation. The red (boron) and blue (gadolinium) dashed circles indicate similar levels of polymerization. Figure 3-5: Comparison of the PCDA/PVA samples containing boron and gadolinium nanoparticles after gamma irradiation. The red (boron) and blue (gadolinium) dashed circles indicate similar levels of polymerization.

34 22 From this experiment it was observed that the samples containing the binding material PMMA polymerized more quickly than the samples containing the binding material PVA. When comparing the different converter materials, it was also observed that samples containing gadolinium nanoparticles polymerized significantly faster than samples containing boron nanoparticles. This was true for the samples that contained the PMMA binder and the samples that contained the PVA binder. The increased sensitivity seen in the samples containing the gadolinium converter is likely due to the higher atomic weight of the gadolinium atom. When gamma rays interact with an atom they can knock out an electron, which in turn can impart its energy to the radiosensitive material causing it to polymerize. Gadolinium atoms contain more electrons than boron atoms, which gives them a higher probability to interact with gamma radiation. 3.3 Neutron Beam Experiments After the samples were tested in a gamma flux, the radiosensitive materials were tested in a thermal neutron flux. Neutrons from Beam Port 4, a well collimated thermal neutron beam facility at the Breazeale Nuclear Reactor, were used to irradiate the samples. For these experiments, the reactor was operated at 900 kw. This corresponded to a thermal neutron flux of approximately 2*10 7 neutrons/cm 2 /s, and a gamma dose rate of approximately 35 rad/hr in the beam. From the gamma ray experiments, it was known that the most limiting sample would not begin to polymerize until it received a gamma ray dose of 60 rad. As long as samples didn t stay in the neutron beam for longer than 1 hour and 40 minutes, any color change seen on the samples would be primarily due to neutron radiation.

35 23 The samples were mounted on paper and taped to a piece of cardboard using kapton tape. These materials were used due to their low neutron cross section and their low level of activation when exposed to neutron radiation. Figure 3-6 shows the sample holder used for the neutron experiments. Figure 3-6: Photograph of the sample holder used in the neutron irradiation experiments. Samples were mounted on paper or silicon wafers then taped to a piece of cardboard with kapton tape. The first neuron experiment tested 16 samples that contained the radiosensitive material PCDA, a binding material of either PVA or PMMA, and nanoparticles of gadolinium or boron. Control samples, which did not contain a converter material, were also created for comparison. The samples were relatively thick and had an approximate thickness between 100 and 200 microns. It is important to note that these sample thicknesses are not the thicknesses envisioned for the final design. The samples thicknesses were chosen to ensure that the exposure time in the neutron beam could be kept under two hours. For this experiment, the samples were inserted into the neutron beam on a transfer track and irradiated for a set amount of time. After each exposure step, the samples were removed from the beam and photographed. Table 3-2 shows the irradiations that were performed on the samples and the equivalent dose that each sample received.

36 24 Table 3-2 The exposures performed on each sample during the first neutron irradiation experiment. Each exposure time is listed, along with the associated neutron dose and gamma dose. Exposure Exposure Total Exposure Neutron Total Neutron Gamma Total Gamma Number Time (min) Time (min) Dose (rad) Dose (rad) Dose (rad) Dose (rad) A number of photographs were taken of the samples during the first experiment after they received various neutron and gamma ray doses in the neutron beam. Some of the photographs were not clear enough for easy comparison. This was due to the activation of the samples in the neutron beam and the resulting required standoff distances, which prevented the photographer from getting close to the samples. Comparisons were able to be made with the PCDA/PMMA samples since they were large in size and easy to photograph from a distance. Direct comparisons were made of the PCDA/PMMA samples containing boron and gadolinium nanoparticles and the photographs are presented in Figure 3-7.

37 25 Figure 3-7: Comparison of the PCDA/PMMA samples containing gadolinium and boron nanoparticles, irradiated during the first neutron experiment. The experiment produced some results that were not expected. Originally it was expected that the samples containing boron would polymerize more quickly than samples that contained gadolinium. This was expected because boron atoms release more energy than gadolinium when they interact with a neutron, which in turn can impart more energy into the radiosensitive material causing it to polymerize. A faster rate of color change was not observed in the samples containing boron, rather it appeared from the photographs that the samples containing gadolinium and the samples containing boron polymerized at approximately the same rate. The research team postulated that this was due to the penetration depth of the particles that are generated from a neutron interaction with the converter material. The internal conversion electrons that come out of a neutron-gadolinium interaction can travel, at a maximum, approximately 80 microns in the radiosensitive material PCDA before they are absorbed. The heavy charged particles that come

38 26 out of a neutron-boron interaction typically travel between five and ten microns in PCDA before they are fully absorbed. Additionally, samples containing gadolinium appeared to polymerize more uniformly than samples containing boron. This seemed to indicate that the increase in energy released from a neutron-boron interaction does not improve the performance of the neutron detector. It was theorized that the penetration depth of the ions generated from a neutron-converter interaction and the thickness of the radiosensitive material, are both key factors in the uniform polymerization of the radiosensitive material. If these two key factors are important for uniform polymerization, increasing the thickness of the radiosensitive material could be detrimental for the samples containing boron. The thick samples containing boron would only cause localized polymerization, but the thick samples containing gadolinium would polymerize the radiosensitive material more uniformly. For the case were the neutron converter is distributed within the radiosensitive material, the concentration density of the boron particles would also be a consideration. Figure 3-8 illustrates what was believed to have happened in the thicker samples. B samples Gd samples 1 n 1 n not polymerized polymerized Figure #3-8: Illustration of how sample thickness affects polymerization. The blue circles represent the area of energy deposition by the emitted particles. This effect was observed in the samples irradiated during the first neutron experiment. A second neutron experiment was performed in the neutron beam with thinner samples to observe how sample thickness affects their sensitivity. These thinner samples were approximately 10 to

39 27 20 microns thick. For this experiment, the samples were deposited onto silicon wafers that were then taped to a piece of cardboard using Kapton tape. The second experiment was again performed with the reactor operating at 900kW, with corresponding neutron flux of approximately 2*10 7 neutrons/cm 2 /s and a gamma ray dose rate of 35 rad/hr. The same irradiation times used in the first neutron experiment, as listed in Table 3-2, were used in the second neutron experiment to give the same neutron and gamma ray dose to the samples. Since some of the photographs did not come out clear enough for easy comparison in the first neutron experiment, only the PCDA/PMMA sample photographs from the second neutron experiment are presented in this section. Direct comparisons were made of the PCDA/PMMA samples that contained boron nanoparticles and the samples that contained gadolinium nanoparticles. The photographs that were taken during the second neutron irradiation are presented in Figure 3-9. Figure #3-9: A comparison of PCDA/PMMA samples containing gadolinium and boron nanoparticles, irradiated during the second neutron experiment.

40 28 The results of the second neutron experiment were expected. Just as in the first experiment, the original expectation was that the samples containing boron would polymerize more quickly than samples that contained gadolinium. It can be observed in Figure 3-9 that the samples containing boron have a greater relative change in polymerization (a darker blue color was observed) than the samples containing gadolinium. This greater relative change is due to the neutron-boron interaction releasing more energy than the neutron-gadolinium interaction. Additionally, polymerization is reduced in the neutron-gadolinium interactions because the emitted electrons can easily escape the thin samples before depositing all of their energy. It was also anticipated that these thinner samples would not polymerize as quickly as the thicker samples that were irradiated in the first neutron experiment. The intensity equation, Equation 3-1, can be used for thin targets to determine how target thickness can affect neutron absorption. [11] This equation gives the uncollided intensity I of the original neutron beam I o after it has traveled through a distance x of the target material with a macroscopic cross section of Σ. = 3-1 From Equation 3-1, it can be observed that as the target thickness decreases, there are less neutron interactions in the target material. The samples from the second irradiation appear to have the same degree of polymerization after approximately 60 minutes of exposure than the samples from the first neutron irradiation had after only 5-10 minutes of exposure. These results from the second test were in agreement with the results from the first test, because the thin samples (10-20 microns) were approximately 5-10 times thinner than the thick samples ( microns). From the neutron irradiations that were performed in the neutron beam, it was observed that the thickness of the radiosensitive material had a significant impact on the uniformity of polymerization throughout the entire sample thickness. This is due to the penetration depth of the ions, emitted from a

41 29 neutron-converter interaction, in the radiosensitive material. It was concluded that to fully optimize the energy released from a neutron-converter interaction, the thickness of the radiosensitive material should be approximately the same as the penetration depth of the ions released from the neutronconverter interaction. The second neutron irradiation indicated that the thinner samples, with nanoparticles of the converter incorporated into the radiosensitive material, were very insensitive to neutron radiation. It was concluded that the Type I detector design, a radiosensitive material mixed with converter nanoparticles, was not sensitive enough for the final detector design. After the neutron experiments, the focus of the research shifted to Type II detector designs that have a layer of the radiosensitive material deposited on a discrete layer of converter material. Modeling of various Type II detector configurations with various converter materials is discussed in more detail in Chapter 4. Penetration depths of the neutron interaction products from various converters, used for the determination of the thickness of the radiosensitive material, are also discussed in more detail in Chapter 4.

42 30 Chapter 4 Modeling of Converter and Radiosensitive Materials The following sections discuss various modeling that was performed in support of the development of the detector. A number of models were run in the Monte Carlo simulation CASINO to analyze the gadolinium metal converter. The Monte Carlo simulation SRIM was used to analyze the converter materials boron nitride and lithium silicate. Both CASINO and SRIM were used to model and analyze how the ions emitted from the converters interact and travel through the radiosensitive material PCDA Neutron Absorption in Gadolinium For this project, it was necessary to gain a better understanding of how much energy could be deposited into the radiosensitive material. Although it was known that not all of the electrons generated from the neutron-gadolinium interactions will reach the radiosensitive material, modeling was performed using CASINO to get an approximation of how much energy can be deposited. Initial models assumed that all neutrons entering the gadolinium metal were at thermal energies (i.e ev), and all electrons were generated with 72 kev of energy. Before modeling was performed in CASINO, the thermal neutron absorption characteristics of gadolinium were quantified. To plot the thermal neutron absorption of gadolinium versus the gadolinium metal depth, the author started with the intensity Equation 3-1 in Section 3.3: = 3-1

43 31 The fraction of neutrons not attenuated by the target of x thickness can then be calculated: = 4-1 be calculated: Finally, the fraction of neutrons attenuated from the beam in the target with a thickness of x can 1 = Since the gadolinium neutron absorption cross section is dominant at thermal energies, and the possibility of multiple reactions within a thin foil is very small, it is reasonable to estimate the number of neutron absorptions in gadolinium using Equation 4-2. Figure 4-1 shows the fraction of neutrons absorbed in the natural gadolinium foil versus the thickness of the gadolinium metal foil. Figure 4-1: The fraction of neutrons absorbed in natural gadolinium versus the gadolinium thickness. Equation 4-2 was used to calculate the fraction of neutrons absorbed in natural gadolinium. It can be observed from this plot that approximately 97% of the thermal neutrons are absorbed in 25 microns of gadolinium. Although the probability of interaction remains constant, fewer and fewer

44 32 neutrons reach the deeper layers of the gadolinium metal foil. To get a realistic approximation of electrons exiting a sample of gadolinium metal foil, the diminished number of neutron absorptions that occur in deeper layers of the gadolinium foil should be taken into account Modeling Methodology The next step in modeling was to understand how the generated electrons interact as they pass through the gadolinium metal, and what fraction of the generated electrons is transmitted out of a given thickness. CASINO was used to model the path of electrons, as well as their spatial energy distribution and the spread of the electrons as they travel through the gadolinium metal converter. One of the computer simulations that was used for modeling electron transport in the detector material was CASINO, or monte CArlo SImulation of electron trajectory in solids. This is a Monte Carlo simulation of a monoenergetic electron s trajectory in a solid material, specifically designed for low beam interaction in a bulk and thin foil. [12] This computer program uses the Monte Carlo Method to model the trajectories and interactions that an electron undergoes when traveling through a certain material. The Monte Carlo Method is a statistical method which solves a problem by generating suitable random numbers and observing that fraction of the numbers obeying some property or properties. This method is useful for obtaining numerical solutions to problems which are too complicated to solve analytically. [13] In particular, the method is useful for solving mathematical and/or physical problems that involve random processes. For this research study, many simulations were performed using a Monte Carlo based computer program, and the calculated result was taken as an average over the number of observations.

45 33 For all the CASINO simulations run in this study, a number of variables were set to the same value. All simulations were performed using CASINO version The number of electrons that were simulated was set to for each CASINO run that was performed. Each CASINO simulation also had the tilt of the specimen set to 0 degrees, and the beam radius set to 0.1 nm. All values in the distribution, option, and physics model pages were kept at the default and were not changed. Values that were modified in the CASINO program included the starting energy of the electron, the composition of the target material, and the thickness of the target material. The results of the CASINO runs are discussed in the following sections. To get a more realistic result, the gadolinium metal converter was broken into one micron thick layers. It was then assumed that all neutrons interactions that occurred in a specific layer, occurred in the center of that layer. Similarly, all electrons born in a specific layer, were assumed to be born in the center of the layer. Each electron that was generated was assumed to have had a 50% chance of being emitted toward the neutron source and a 50% chance of being emitted away from the neutron source. The model then coupled the probability that a neutron interacts in a certain layer, with the probability that the resulting electron can escape a given thickness of gadolinium. The fraction of energy that the resulting electrons retains, if it escapes the gadolinium, was also factored into the model. For this preliminary analysis, all neutrons were assumed to be at thermal energies, and all electrons that were born were assumed to have an initial energy of 72 kev. Figure 4-2 below shows an illustration of how the gadolinium metal was broken into one micron layers for the modeling.

46 34 Figure 4-2: Illustration of modeling methodology used in the gadolinium metal analysis. The converter material was broken into 1 micron layers, and it was assumed that all neutron-converter interactions occurred in the center of each layer. Analysis was performed with the gadolinium metal foil in two different configurations. The first configuration placed the gadolinium metal foil in between the radiosensitive material and the neutron source. The second configuration placed the radiosensitive material in between the gadolinium foil and the neutron source. Figure 4-3 and Figure 4-4 illustrate the two modeling configurations that were analyzed.

47 35 Figure 4-3: Illustration of the first detector configuration with the converter material placed between the radiosensitive material and the neutron source. Figure 4-4: Illustration of the second detector configuration with the radiosensitive material between the converter material and the neutron source.

48 Gadolinium Metal Analysis with Average Energy Electrons To determine the optimum thickness of gadolinium foil, where the energy emitted from the foil is highest, the foil was analyzed using CASINO. The fraction of electron energy that can be emitted from the gadolinium foil was calculated for the first configuration with gadolinium metal placed in front of the radiosensitive material, as illustrated in Figure 4-3. A common thickness for commercially available gadolinium foil is 25 microns, so the following analysis was performed for gadolinium thicknesses up to 25 microns. The graph presented in Figure 4-5 shows the fraction of electron energy that can be transferred through the gadolinium foil for increasing foil thickness. Figure 4-5: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the first configuration versus the gadolinium metal foil thickness. The electrons are emitted from the gadolinium converter with an average energy of 72 kev. From Figure 4-5, it can be observed that the maximum amount energy that can be emitted from the gadolinium foil, oriented in the first detector configuration, occurs around a thickness of 5 microns. For thicknesses thinner than 5 micron, the fraction of neutrons that get absorbed by gadolinium

49 37 decreases, which reduces the number of generated conversion electrons, and thus reduces the energy that can be transferred to the radiosensitive material. As the foil thickness increases beyond 5 microns the fraction of incident neutrons that are absorbed by gadolinium atoms increases, but electron attenuation prevents electrons generated in the front layers of the foil (small values of x) from escaping the foil. With more gadolinium atoms present in thicker foils, the electrons have to travel through more material and lose all of their energy before they can exit the foil. For the second configuration, where the gadolinium foil is placed behind the radiosensitive material as illustrated in Figure 4-4, the fraction of electron energy that can be emitted from the gadolinium foil into the radiosensitive material was calculated. Again, the following analysis was performed for gadolinium up to thickness of 25 microns. Figure 4-6 shows the fraction of electron energy that can be emitted from the gadolinium foil versus an increasing gadolinium metal foil thickness. Figure 4-6: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the second configuration versus the gadolinium metal foil thickness. The electrons are emitted from the gadolinium converter with an average energy of 72 kev.

50 38 From Figure 4-6, it can be observed that the maximum fraction of electron energy that can be emitted from the gadolinium foil, oriented in the first detector configuration, occurs for thicknesses greater than approximately 7 microns. For thicknesses thinner than approximately 7 micron the fraction of neutrons that get absorbed by gadolinium decreases, which reduces the energy that can be emitted from the foil. As the foil thickness increases beyond 7 microns, more neutrons get absorbed by the gadolinium foil but the electron energy that gets emitted from the gadolinium foil plateaus. This occurs because the electrons generated in the deeper depths of the gadolinium foil lose all of their energy before they can escape. The electrons that are generated in the shallower layers do not have to travel through addition material to escape in the direction of the radiosensitive material as the foil thickness increases. It is important to note that in this configuration there is no downside for increasing the thickness of the gadolinium foil as long as the backward configuration is maintained Average Energy Electrons versus Multiple Energy Electrons As the research progressed, a concern was raised as to whether it was reasonable to assume that all electrons are born in the gadolinium foil with an energy of 72 kev. To validate this approximation, additional modeling was performed that allowed the electrons to be born at various energies. Table 2-5, which shows the most frequent energies that electrons can be born with during a neutron gadolinium interaction, is presented again as Table 4-1. The energy groups that were used in the analysis in this section are highlighted in yellow.

51 39 Table 4-1 Energy and frequency of emission for internal conversion electrons released from neutron gadolinium interactions. These internal conversion electrons are the primary contributors to neutron radiography. Entries that were used in the analysis are highlighted in yellow. Energy Group Energy Frequency Range in Gd [CSDA] Energy/Interaction % of Energy (kev) (%) (μm) (kev) (%) total: For this analysis, the same method was used to model the gadolinium foil. The gadolinium metal was broken into 1 micron layers. It was then assumed that all neutrons interactions that occurred in a specific layer, occurred in the center of that layer. Similarly, all electrons born in a specific layer, were assumed to be born in the center of the layer and had a 50% chance of being emitted in the direction of the radiosensitive material. The angular distribution of electron emission was not taken into account, as all electrons were assumed to be emitted either backward or forward along a line perpendicular to the gadolinium metal foil surface. The model then coupled the probability that a neutron interacts in a certain layer, with the probability that the resulting electron can escape a given thickness of gadolinium. The fraction of energy that the resulting electrons retain, if it escapes the gadolinium, was also factored into the model. Again, all neutrons were assumed to be at thermal energies.

52 40 The detector configurations illustrated in Figure 4-3 and Figure 4-4 where once again used in this analysis. CASINO was used to model electrons at each energy level as they traveled through various thicknesses of gadolinium metal foil. The graph in Figure 4-7 shows the fraction of electron energy emitted from the gadolinium metal foil versus gadolinium thickness, for the first detector configuration. For comparison, the average electron energy data analysis shown in Section 4.3 was plotted with the multi-energy electron data. Figure 4-7: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the first configuration versus the gadolinium metal foil thickness. The average energy data is represented by blue diamonds and the multiple energy data is represented by red squares. From Figure 4-7, it can be observed that for the multi-energy data, the maximum fraction of electron energy emitted from the gadolinium metal foil, oriented in the first detector configuration, occurs at a thickness of 5-6 microns. The peak from the multi-energy analysis occurs at approximately the same gadolinium thickness as the average energy analysis, but the magnitude is slightly smaller. It can also be observed that for increasing gadolinium thicknesses, the fraction of energy emitted from the gadolinium does not decrease as quickly for the multi-energy data as it does for the average energy

53 41 data. This is due to the high energy electrons that are modeled in the multi-energy analysis, which have a much higher probability to penetrate the thicker layers of gadolinium. Figure 4-8 shows the results for the second detector configuration, where the radiosensitive material is placed between the neutron source and the gadolinium metal foil. The fraction of electron energy emitted from the gadolinium foil, is plotted against the gadolinium metal foil thickness. Again, the data from Section 4.3 is plotted on the figure for comparison. Figure 4-8: The fraction of energy emitted from the gadolinium converter layer of a detector oriented in the second configuration versus the gadolinium metal foil thickness. The average energy data is represented by blue diamonds and the multiple energy data is represented by red squares. From Figure 4-8, it can be observed that the multi energy data and the average energy data reach approximately the same plateau of the fraction of energy emitted from the gadolinium foil, oriented in the second detector configuration. As the gadolinium metal foil thickness increases, the fraction of energy emitted from the gadolinium foil increases slower for the multi-energy data than it does for the average energy data. While both sets of data reach the same fraction of energy emitted

54 42 plateau, the multi-energy data does not reach the plateau until a gadolinium thickness of approximately 10 microns. When comparing the two detector configurations, it is observed that the second configuration reaches a higher fraction of electron energy emitted from the gadolinium. Again, it is important to mention that there is no downside to increasing the thickness of the gadolinium metal in this configuration, as long as the orientation with respect to the neutron source is maintained. This analysis shows that it is reasonable to assume that all electrons created during a neutron gadolinium interaction can be assumed to have an average energy. To ensure full agreement in the amount of energy emitted between the average energy and the multi-energy models, the second detector configuration should be used. Again, the second detector configuration places the radiosensitive material between the neutron source and the gadolinium metal foil. The detector should also have a minimum gadolinium metal foil thickness of 10 microns Neutron-Gadolinium Interaction Products in PCDA In support of the development of the radiosensitive material in the detector, analysis was performed on the radiosensitive material PCDA. An important design consideration for the detector is determining how thick the radiosensitive material should be in the device. If the radiosensitive material is too thin, very little of the energy emitted from the converter will be absorbed in the radiosensitive material. If the radiosensitive material is too thick, the radiosensitive material will absorb the majority of energy emitted from the converter but the energy will not be evenly distributed throughout the entire thickness of the radiosensitive material. To determine how thick the radiosensitive material should be, Monte Carlo simulations were performed on the charged particles that are emitted from neutron interactions with various converter materials.

55 43 The Monte Carlo simulation CASINO was used to model the path of electron, created during a neutron gadolinium interaction, as it traveled through the radiosensitive material PCDA. The multienergy analysis that was discussed in Section 4.4 showed that it is reasonable to assume that all electrons created during a neutron-gadolinium interaction have an average energy of 72 kev. Figure 4-9 shows the CASINO tracks of 72 kev electrons traveling through the radiosensitive material PCDA. Figure 4-9: The electron track plot generated in CASINO, for 72 kev electrons in PCDA. To get a better understanding of where the electrons deposit their energy in the radiosensitive material, a different CASINO plot needs to be analyzed. Figure 4-10 shows the electron energy deposition distribution in PCDA, for 72 kev electrons, for the electron tracks presented in Figure 4-9. In this plot, 95% of the electron energy is contained within the bluish-green line. Measurements of the depth and width of the energy deposition distribution are also shown on the figure.

56 44 Figure 4-10: The electron distribution plot generated in CASINO, for 72 kev electrons in PCDA. From Figure 4-10, it can be observed that the electrons deposit their energy in an area that is approximately 80 microns by 80 microns in PCDA. The wide distribution of electrons in the target material was expected, since the incident electrons are significantly scattered rather than penetrate the target in a linear fashion. [14] This scenario accurately represents electrons that are born on the surface of the gadolinium metal foil. However, electrons can be born at various depths of the gadolinium foil not just at the surface. Approximately 50% of the incident neutrons will interact within the first five microns of the gadolinium metal. Before the resultant electrons can reach the radiosensitive material, they will lose a portion of their energy in the gadolinium metal foil. The optimum thickness of the radiosensitive material should be set to a value that is less than 80 microns, to allow the radiosensitive material to polymerize more uniformly.