Design and Development of an External Fast Neutron Beam Facility at the Ohio State University Research Reactor

Thesis

By Andrew M. Zapp Graduate Program in Nuclear Engineering

The Ohio State University 2019

Master‟s Examination Committee: Dr. Vaibhav Sinha, Advisor Dr. Lei Cao

Abstract

The ability of the Ohio State University Research Reactor, OSURR, to conduct experiments from the generation of the neutron flux is important in conducting research by the

University and external entities that require a flux of this magnitude. In particular, research involving a fast neutron flux is of interest due to the different interactions fast neutrons have as opposed to thermal neutrons. The OSURR is able to operate up to 500 kW, which creates a neutron flux in the order of 1013 n/cm2-s. Currently, Beam Port 2 provides a thermal neutron beam profile of 30 mm in diameter for experimentation such as neutron depth profiling, activation analysis, and evaluation of radiation damage to electronics. Beam Port 1 uses a sample area located adjacent and perpendicular to the fuel plates of the reactor for in-core irradiation. During experimentation, the remainder of Beam Port 1 must be plugged with removable concrete shielding to prevent radiation exposure that can be upwards of 1x104 rem/hr.

The upgrade to Beam Port 1 consists of a collimator to shape the neutron flux from the reactor into a beam of fast neutrons, similar in diameter to Beam Port 2, in order to irradiate samples external to the reactor. In addition, mobile external shielding is designed to prevent exceeding the exposure limits of 5 rem/yr when the facility is in use. With this upgrade the research reactor has the ability to conduct simultaneous experiments with a fast and thermal neutron beam, external to the biological shielding, without releasing any harmful exposure.

Acknowledgements

I would first like to thank my Nuclear Engineering graduate advisor, Dr. Vaibhav Sinha, for his help throughout this process.

I would also like to thank the efforts conducted by my fellow graduate students Ibrahim

Oksuz and Matthew Van Zile, who are in the Ohio State University Nuclear Engineering program, under the supervision of Dr. Lei Cao.

This would not be possible without the substantial support and guidance from the staff at the Ohio State University Nuclear Reactor Laboratory. In particular, Mr. Andrew Kauffman, the

Associate Director, and Mr. Joel Hatch, the Research Engineer, provided a significant amount of their time and effort in collaboration of this design.

This work is supported by The Ohio State University and the Department of Energy through NEUP General Scientific Infrastructure Support Program for grant no. RU-17-13347, “A

Request for Upgrade of the Ohio State University Research Reactor Beam Ports Infrastructure”.

Vita

2005 B.A. International Studies, Russian Minor,

Miami University (OH)

2008 Naval Nuclear Training Command Graduate

2007-2015 Nuclear trained Machinist Mate,

USS Seawolf

2018 Presentation at SORMA XVII, “Design and

Development of an External Fast Neutron

Beam Facility at the Ohio State University

Research Reactor”, University of Michigan

Fields of Study

Major Field: Nuclear Engineering

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Table of Contents

Abstract i Acknowledgements ii Vita iii Fields of Study iii List of Figures vii List of Tables x 1 Introduction 1 2 Theory 4 2.1 Reactor/Source 4 2.1.1 Neutron Interactions/Properties 6 2.1.2 Gamma Interactions/Properties 8 2.2 Shielding 9 2.2.1 Neutron Shielding 9 2.2.2 Gamma Shielding 10 2.3 Collimation 12 2.3.1 Illuminator 16 2.3.2 Beam Filter 17 2.3.3 Aperture 19 3 Materials 20 3.1 Collimator Materials 20 3.1.1 Graphite Illuminator 20 3.1.2 Bismuth and Sapphire Filters 25 3.1.3 Aperture Pieces 34 3.1.3.1 Borated Cement 36 3.1.3.2 Metamic® (Borated Aluminum) 38 3.1.3.3 Lead 40 3.1.4 Aluminum 6061 as Structural Material 42 3.2 Beam Stop Shielding Materials 45

3.2.1 Borated Polyethylene 45 3.2.2 Metamic® 46 3.2.3 Lead 47 4 Design 49 4.1 Existing Structure 51 4.1.1 Beam Port 2 Facility 51 4.1.2 Beam Port 1 Current Lay-out 53 4.2 MCNP Simulations 60 4.2.1 Collimator MCNP Simulations 61 4.2.2 Beam Stop MCNP Simulations 62 4.3 Design Calculations 63 4.3.1 Collimator Calculations 63 4.3.2 Shielding Calculations 67 4.3.2.1 Fast Neutrons 67 4.3.2.2 Gammas 70 4.4 Collimator Final Designs 70 4.4.1 Design Process 70 4.4.2 Beam Port 1 72 4.4.2.1 Inner Collimator 72 4.4.2.2 Outer Collimator 74 4.4.3 Beam Port 2 75 4.4.3.1 Inner Collimator 75 4.4.4 Aluminum Structural Considerations 77 4.5 External Shielding Design 80 4.5.1 Beam Stop 81 4.5.2 Rail/Lift System 84 4.5.2.1 PBC Linear Components 85 4.5.2.2 Joyce-Dayton Components 87 4.5.2.3 Fabricated Components 90 4.6 Future Work 91 5 Conclusions 93

References 95

Appendix A: Drawings Index 99

Appendix B: MATLAB Code for Table 4-6 103

Appendix C: MCNP Code for Figure 4-9 104

Appendix D: Additional Weight and Cost Estimates 110

List of Figures

Figure 2-1: Elastic neutron scattering collision 6 Figure 2-2: Different gamma interactions with respect to the materials Z number and gamma energy (Knoll 2010) 9 Figure 2-3: Attenuation coefficients based on energy for NaI (Knoll 2010) 12 Figure 2-4: General image plane for a source showing the image plane and distance (ASTM E803 1996) 14 Figure 2-5: Final Collimator design for the Beam Port 2 upgrade in 2012 (Turkoglu 2012) 15 Figure 2-6: Neutron Cross Sections for Carbon (Lamarsh 2001) 17 Figure 3-1: Graphite, Beam Port 1 23 Figure 3-2: Graphite, Beam Port 2 24 Figure 3-3: 3-D image of Graphite with aluminum case for Beam Port 2 25 Figure 3-4: Three single-crystal Sapphires (left) used in the 2012 upgrade to Beam Port 2 (Turklogu 2012) 26 Figure 3-5: Gamma attenuation for different thicknesses of Bismuth 27 Figure 3-6: Bismuth total cross-section for 1 - 600 meV neutron energy ranges (Freund 1983) 28 Figure 3-7: Bismuth total cross-section for 10-3 - 10 eV neutron energy ranges (Adib 2003) 29 Figure 3-8: Total neutron cross section for 209Bi, from .0125 eV – 5 MeV (NNDC) 30 Figure 3-9: Poly-crystal Bismuth molded into an aluminum holder (Turklogu 2012) 31 Figure 3-10: 3-D image of new Sapphire holder 32 Figure 3-11: Bismuth encased in Aluminum shell 33 3 6 10 Figure 3-12: σabs of He, Li, and B for a large range of neutron energies (Knoll 2010) 35 Figure 3-13: Borated Cement piece for Inner Collimator 38 Figure 3-14: Metamic® Apertures 40 Figure 3-15: Attenuation coefficients for Lead (Lamarsh 2001) 41

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Figure 3-16: Lead Apertures 41 Figure 3-17: Stress-Strain Curves (Oberg 2012) 42 Figure 3-18: Example of Borated Polyethylene sheet drawing 46 Figure 3-19: Example of Metamic® sheet drawing 47 Figure 3-20: 3-D model of lead sheet 48 Figure 4-1: Rear View of Beam Port 2 (OSURR n.d.) 52 Figure 4-2: Overhead view of Beam Port 2 (OSURR n.d.) 53 Figure 4-3: Beam Port 1 at exit of the port (OSURR n.d.) 54 Figure 4-4: 3-D overhead view of Beam Ports inside the OSURR 55 Figure 4-5: 3-D view of Beam Ports inside the OSURR 56 Figure 4-6: 3-D cut-out of the Shutter Box Assembly in the “Open” (left) and “Shut” (right) position 57 Figure 4-7: Removal of the Shutter Box Assembly (OSURR n.d.) 57 Figure 4-8: Shutter Box Assembly when removed for maintenance (OSURR n.d.) 58 Figure 4-9: MCNP model of the collimator 60 Figure 4-10: Neutron transmittance through 4” of Bismuth from neutron energies of .0125 eV – 10 MeV 64 Figure 4-11: Neutron transmittance showing the energy range corresponding to 1/v cross section range (.0125 eV – 1 keV) 65 Figure 4-12: Final neutron flux through Bismuth 66 Figure 4-13: Collimated fast neutron flux 66 Figure 4-14: Example of drawing assembly 71 Figure 4-15: Example of individual drawing used for above assembly 72 Figure 4-16: Inner Collimator, Beam Port 1 73 Figure 4-17: Inner Collimator Aluminum Shell, Beam Port 1 73 Figure 4-18: Outer Collimator 74 Figure 4-19: Inner Collimator Aluminum Shell 75 Figure 4-20: Inner Collimator, Beam Port 2 76 Figure 4-21: Inner Collimator Aluminum Shell, Beam Port 2 76 Figure 4-22: Shield plug handling tool end (OSURR n.d.) 77 Figure 4-23: 3-D model of Aluminum shell cap design for removing the collimator 78

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Figure 4-24: Stress and beam load case for Aluminum shell extending into the reactor pool (Oberg 2012) 79 Figure 4-25: 3-D Model of Beam Stop 82 Figure 4-26: 3-D model of Beam Stop cut-out showing Lead inserted in the back 83 Figure 4-27: Double roller pillow block data from PBC Linear (PBC Linear n.d.) 85 Figure 4-28: 3-D model of PBC linear components with bottom frame 86 Figure 4-29: 3-D model showing the removable section of railing 86 Figure 4-30: Beam Port 2 (Left) Beam Shutter without Lead components, (Right) 3-D model of Beam Shutter with Lead components (Turklogu 2012) 87 Figure 4-31: Typical machine screw jack with a lifting screw (Joyce/Dayton n.d.) 88 Figure 4-32: Machine screw jack with a traveling nut (Joyce/Dayton n.d.) 89 Figure 4-33: 3-D model of Joyce-Dayton Components 90 Figure 4-34: 3-D model of bottom frame (left) and middle frame (right) 91 Figure 4-35: 3-D Model of all components 91

List of Tables

Table 2-1: OSURR Neutron flux data at different areas of the core (OSURR n.d.) 4 Table 2-2: Energy loss values for Hydrogen and Carbon 7 Table 3-1: Carbon nuclide data (NNDC n.d.) 21 Table 3-2: Comparison of the three filter materials and associated densities and cross sections 30 Table 3-3: Bismuth nuclide data (NNDC n.d.) 31 Table 3-4: Boron nuclide data (NNDC n.d.) 34 Table 3-5: Composition data of Borated Cement vs. Borated Polyethylene (Shieldwerx n.d.) 37 Table 3-6: Metamic® composition data from Reynolds Metal Company (Holtec Int. n.d.) 39 Table 3-7: Aluminum nuclide data (Mondolfo 1976) 43

Table 3-8: Thermal Neutron σabs for Several Metals (Tsoulfanidis 2013) 43

Table 3-9: Al-6061 composition with Mg2Si (ASM n.d.) 44 Table 4-1: Collimation ratio 64

Table 4-2: ΣR for Borated Polyethylene 68

Table 4-3: ΣR for Metamic® 68

Table 4-4: ΣR for Lead 68 Table 4-5: Fast neutron HVL for shielding components 69 Table 4-6: Shielding thickness for fast neutrons 69 Table 4-7: Gamma dose due to neutron capture reactions 70 Table 4-8: Summary of material thicknesses, Inner Collimator, Beam Port 1 73 Table 4-9: Summary of material thicknesses, Outer Collimator 74 Table 4-10: Summary of material thicknesses, Inner Collimator, Beam Port 2 76 Table 4-11: Max stress and deflection calculations 80 Table 4-12: Beam Stop estimated weight 84 Table 4-13: Lifting Screw Data 89 Table 4-14: Lifting Screw Performance/Safety Details 89 Table A-1: Drawing Index 99

Table D-1: Beam Port 1, Inner Collimator estimated weight 110 Table D-2: Beam Port 2, Inner Collimator estimated weight 110 Table D-3: Outer Collimator estimated weight 111 Table D-4: Rail/Lift system estimated cost excluding fabricated steel pieces 111

1. Introduction

The Ohio State University Research Reactor (OSURR) was completed in 1960 as a pool type design from Lockheed Nuclear Products. It first went critical in 1961 using highly enriched uranium solid plate fuel, a natural convection cooling method, and a max power of 10 kW.

Originally used as a training reactor, the purpose of the reactor at present is for research, and is the only research reactor in the state of Ohio. After going through several improvements since that time, it is now licensed to operate up to 500 kW by utilizing low enriched uranium solid fuel plates. The intention of this project is to make additional improvements by upgrading the beam facilities in order to increase the research capabilities of the OSURR.

The research is accomplished with use of several irradiation facilities in the reactor that, at full power, are capable of being exposed to neutron fluxes in the range of 1012-1013 n/cm2/s.

These facilities include three dry tubes, two beam ports, a “rabbit” pneumatic system, and a thermal column. The beam ports are radial to the reactor, carved cylindrically into the biological shielding of the reactor with a diameter of 6”, and have aluminum extensions passing through the reactor pool facing the fuel plates of the core. Geometrically, Beam Port 1 is perpendicular to the reactor core while Beam Port 2 extends from the fuel plates at a 30 degree angle.

Originally, these ports were used to insert research samples for in-core or near core irradiation. This process would allow a sample to be exposed to the isotropic neutron and gamma flux of the reactor for a given time and power of reactor operation. This, however, limited research applications to mainly neutron activation analysis and radiation damage

1 evaluation. In 2012, Beam Port 2 underwent an upgrade that allowed a sample to be exposed to a thermal neutron flux of ~106 n/cm2/s, 30 mm in diameter, external to the biological shielding of the reactor. This upgrade expanded the capabilities of the research reactor by moving the irradiation outside of the biological shielding. Not only did this allow for more research opportunities such as neutron imaging and neutron depth profiling, but by using a beam of neutrons, the angle of exposure is able to be included into the sampling process. This upgrade also required a significant amount of external shielding due to the creation of a neutron beam outside of the biological shielding. At full power, exposure from the thermal neutron beam, surrounding the facility, is reduced to less than the limit of 5 Rem/year.

This project is mainly an upgrade to Beam Port 1, but also includes improvements to the design of the upgrade for Beam Port 2 in 2012. The upgrade to Beam Port 1 allows a sample to be exposed to a beam of fast neutrons, 1.25” in diameter, external to the biological shielding of the reactor. This upgrade also includes external shielding that allows the fast neutron beam passing through the sample to reduce exposure to personnel around the facility to less than the radiological limits required by the NRC. The external shielding is designed on rails for horizontal movement, and a lift system for vertical movement. Lastly, an automation feature is added to the existing beam port shutter assembly to allow the shutter to be used as additional shielding when the facility is not in use.

The improvements to Beam Port 2 focus mainly on replacing the majority of the components in the collimator. One reason for this change is to allow more uniformity in design between the two Beam Port facilities, and so that both are capable for any kind of modulation in the future.

Secondly, this new collimator design improves upon the majority of the existing materials used in Beam Port 2 to increase the longevity of the collimator over time as it is exposed to high

2 energy gammas and neutrons. The final design of both collimators is similar, with the exception of the filters used to shape either a fast or thermal neutron beam.

This project allows the OSURR to have two external neutron beam facilities, one fast and one thermal, with a uniform fabrication process to allow modulation for any changes in the future.

2. Theory

2.1 Reactor/Source

The OSURR is able to operate at a max power of 500 kW, and from this can produce a neutron flux in the order of 1012-1013 n/cm2/s in the beam facility area. While neutrons are born as fast neutrons, the neutron spectrum ranges in energy from thermal energies, < 1eV, up to several MeV.

Neutrons, however, are not the only high energy particle produced from a fission reactor.

High energy gammas are also created either directly from fission, or from secondary interactions with neutrons. The fission reaction process produces a neutron and gamma source that is isotropic in nature. Prompt neutrons comprise more than 99% of the neutrons in these reactions and are born as fast neutrons with an average energy of 2 MeV (Lamarsh 2001).

Table 2-1: OSURR Neutron flux data at different areas of the core (OSURR n.d.).

Neutro Gamma Total Percent Thermal epi-Cd 1.0 MeV Eq n Dose Dose in Neutron Thermal Neutron Neutron Neutron Rate in Si Facility Flux Flux Flux Flux Si

(n/cm2/s) (%) (n/cm2/s) (n/cm2/s) (n/cm2/s) (rad- (rad- Si/hr) Si/hr) FD,8.0 MeV,Si = 95 MeV-mb (E < 0.5 (E > 0.5 n n eV) eV) FD,1.0 MeV,Si = 70 MeV-mb

(Ref: ASTM E722- 14) 12 † CIF 2.3x1013* 59 1.4x1013* 1.0x1013* 4.7x10 (Si) 1.4x106 8.7x107† 5.1x1012 (GaAs)† 12 † AIF 9.4x1012$ 48 4.5x1012$ 4.9x1012$ 2.6x10 (Si) 7.6x105 3.8x107† 2.7x1012 (GaAs)† 12 † PIF 5.3x1012** 57 3.1x1012** 2.3x1012** 1.2x10 (Si) 3.7x105 2.5x107† 1.3x1012 (GaAs)† 11 † Rabbit 3.2x1012†† 65 2.1x1012†† 1.1x1012†† 5.3x10 (Si) 1.6x105 5.9x1011 (GaAs)† 7” 11 † Moveable 1.5x1012†† 73 1.1x1012†† 4.1x1011†† 2.0x10 (Si) 6.5x104 7.6x106† 2.3x1011 (GaAs)† Dry Tube 10” Moveable 11 † Dry Tube 6.1x1011$$ 70 4.3x1011$$ 1.8x1011$$ 1.0x10 (Si) 3.1x104 2.5x106† 1.1x1011 (GaAs)† (Flux Box Evacuated) Beam Port 1.0x1012 (Si)† #1 Sample 12*** 12*** 12*** 5 4.2x10 54 2.3x10 1.9x10 1.1x1012 3.1x10 Holder (GaAs)† Position Beam Port #2 External 4.4x106††† Beam Line Facility Thermal Column G7, core 2.5x1011††† end, stringer open Note: All values are for 450 kW operations, except for the 10” tube which is limited to 250 kW

* NRL memo 2013-09, “CIF Neutron Spectrum Measurent” † NRL memo 2013-12 Rev. 02, “1.0 MeV Equivalent Flux Calculations in Si and GaAs” † NRL memo 2016-17 Rev. 01, “Gamma Dose Rates of Reactor Facilities” $ NRL memo 2013-10, “AIF Neutron Spectrum Measurement” ** NRL memo 2013-11, “PIF Neutron Spectrum Measurement” †† NRL memo 2013-05, “Rabbit Neutron Spectrum Measurement” †† NRL memo 2014-01, “Seven-inch Tube Neutron Spectrum Measurement” $$ NRL memo 2015-06 Rev. 02, “Ten-Inch Dry Tube Neutron Spectrum Measurement” *** NRL memo 2015-07, “BP1 Sample Holder Neutron Spectrum Measurement” ††† NRL memo 2015-05, “BP2 Facility Flux Measurements” ††† NRL memo 2015-08, “TC Thermal Flux Measurements”

2.1.1 Neutron Interactions/Properties

Being electrically neutral, neutrons are not affected by the negative charge of electrons or by the positive charge of the nucleus. This key property allows neutrons to pass through the atomic electron cloud and interact directly with the nucleus. In general, neutron scattering and absorption with a nucleus are the two primary methods of neutron interaction. These two types of interactions can occur in different probabilities depending on the neutron energy. The total neutron cross-section can change over a wide range of energies, and are typically differentiated between fast and thermal neutrons. This can be seen in Figures 3-8 and 3-12.

Scattering interactions can be classified into two categories: elastic and inelastic. Elastic scattering is the most prominent type of neutron scattering where a significant amount of energy is lost. In this type, the neutron strikes a nucleus like a billiard ball, leaving the nucleus in its ground state. Inelastic scattering is similar, except that the nucleus is left in an excited state after a collision. Figure 2-2 shows a typical neutron elastic scattering collision with an atom.

Figure 2-1: Elastic neutron scattering collision.

Where: En = Initial neutron energy MR = Mass of the nucleus (at rest) E‟ = Final neutron energy ER = Final nucleus energy Using the Conservation of Energy and Momentum Laws, and assuming the initial collision atom is at rest, an equation can be found for light nuclei that can find the average energy a neutron will lose per collision.

Avg. energy loss per collision (E‟ ): (2.1) n,avg

Where: ( )

A = Atomic mass From this, a typical moderator such as Graphite can be chosen to best slow down neutrons to thermal energies. Some typical values of α of two moderator materials used for this project are in Table 2-1. The Carbon is utilized in Graphite, and the Hydrogen is an important shielding element that is in Borated Polyethylene and Cement. More detail about these materials, their importance, and more specific uses of Hydrogen and Carbon are discussed later.

Table 2-2: Energy loss values for Hydrogen and Carbon.

Nucleus Mass No. α

Hydrogen 1 0 .5*En

Carbon 12 .716 .358*En

Neutron absorption interactions can result in several different outcomes. These are radiative capture, charged-particle or neutron producing reactions, and fission. One that is common to most nuclei is radiative capture, in which a nucleus absorbs the neutron and in turn emits a gamma. Another that occurs in only a few heavy nuclei is fission, in which the nucleus

7 splits in two after absorbing a neutron, releasing a large amount of energy. These two types of absorption examples are not in themselves of the greatest concern for this upgrade, but rather the absorption cross section, σabs, itself. Regardless of the outcome, the result of an absorption interaction is the loss of a neutron and a different particle being emitted. The specifics of the more significant neutron absorption interactions will be discussed later.

2.1.2 Gamma Interactions/Properties

Gammas, like neutrons, are also electrically neutral, but have different interactions with matter, and since they have no mass, are able to interact with the electron cloud of atoms through scattering. The three processes that define how gammas interact with matter are the

Photoelectric Effect, σpe, Pair Production, σpp, and Compton Scattering, σcs. Photoelectric Effect is when the gamma disappears after it interacts with the entire atom, and occurs with lower energy gamma rays. In Pair Production, the photon becomes an electron pair, and must occur with photon energy of at least 1.02 MeV, which is the rest-mass energy of 2 electrons. Lastly, the Compton Effect, which occurs in the energy range between the other two interactions, is simply the elastic scattering of a photon with an electron. The total cross-section, σtot, for gamma interactions are the sum of these three. The probability of each type of interaction is dependent upon the energy of the incoming gamma ray and has a proportional dependence on the number of protons, Z-value, of the material.

Figure 2-2: Different gamma interactions with respect to the materials Z number and gamma energy (Knoll 2010).

2.2 Shielding

2.2.1 Neutron Shielding

Neutron shielding is accomplished by attenuating neutrons through shielding materials.

Neutron attenuation is determined by the total macroscopic cross section, Σt, of the shielding material which is found from the sum of the absorption and scattering microscopic cross sections, σabs + σs. The Σt is determined by measuring the intensity of a neutron beam traveling through the shielding material. The equation for neutron attenuation is the simple exponential law,

Neutron attenuation (φ): (2.2)

(2.3) -3 Where: Ni = Atomic density of material (cm ) x = Thickness of material (cm) υ(x) = Neutron flux through material of thickness x, (cm-2s-1)

υo = Initial neutron flux

This equation is useful for shielding and radiography calculations. From this, the transmitted neutron can be calculated to determine amount of shielding necessary for neutrons as well as the amount of neutrons passing through an object for radiography. This equation is also used in designing certain aspects of the collimator such as the filter. For fast neutrons, the neutrons are

„removed‟ by scattering to thermal energies and being absorbed. Therefore, in the fast energy ranges, the total microscopic cross section is referred to as the removal cross section, ΣR.

2.2.2 Gamma Shielding

Similar to neutron shielding, gammas are also shielded by attenuation. When multiplying the atomic density of the colliding atoms by these cross sections, the result is called the attenuation coefficients rather than the macroscopic cross-section, and is denoted by the symbol,

μ. Additionally, density plays an important role in attenuating gammas since the denser a material the closer the atoms are to each other, therefore, the higher the probability of gamma interactions. This leads to the mass attenuation coefficient, which is the attenuation coefficient,

μ, divided by a materials density, ρ.

(2.4)

(2.5)

The mass attenuation coefficient can then be used to find gamma attenuation in a specific medium, which is shown in equation. However, since each individual mass attenuation coefficient is energy dependent on the incoming gamma, this changes the total mass attenuation coefficient as a function of gamma ray energy. This means that the colliding material will have a range of mass attenuation coefficient values over the energy range of the gammas from the source. An example of the mass attenuation coefficients as a function of gamma ray energy is shown in Figure 2-4.

Gamma attenuation (I): or (2.6)

(2.7) Where: I(x) = Gamma intensity through material of thickness x

I0 = Initial Gamma intensity An additional aspect of gamma shielding is that they do not disappear as a result of the above interactions, but may produce less energetic x-rays. This leads to the exposure buildup factor, Bm(μx), for a mono-directional beam, and these values are already calculated for several materials.

Gamma attenuation with Buildup (Ib): (2.8)

Where: Bm = Dose-buildup factor

When using the equation for exposure rate, it is also possible to find the gamma exposure rate for a given energy that passes through shielding.

Exposure rate, unshielded: ̇ (2.9)

Exposure rate, shielded: ̇ ̇ (2.10) Where: ̇ = exposure rate (mR/hr)

Eo = Energy in MeV

= .0297 cm2/g

Figure 2-3: Attenuation coefficients based on energy for NaI (Evans 1955).

2.3 Collimation

Collimators are not just limited to research reactors, and many different designs provide outputs that do not produce a parallel beam of neutrons from a reactor. In terms of the shape of the output, collimators can be either convergent, divergent, or in our case parallel. In the medical

12 field, where collimators are more prevalent, a knife-edge type collimator, for example, is used in a gamma camera for medical diagnosis (Wanno 2009). The knife-edge is a pinhole type of collimator that can have an aperture diameter of 4-8mm, and creates a slight convergent beam of photons onto a detector for imaging. The output is based on the design of the aperture component of the collimator, and is discussed below.

For this facility, the collimation process produces a beam of neutrons with directions that are within a few degrees of being parallel (Turklogu 2012). This is achieved by allowing the desired particles moving in the correct direction to travel relatively unimpeded while simultaneously removing all of the unwanted particles. ASTM E803 defines a collimation ration of a neutron radiography beam as the distance between the source and the image plane divided by the diameter of the source. In the simplest of terms, this requires having an adequate collimation ratio (equation below), which is the ratio of the collimator length to effective diameter of the aperture (McGillivray n.d.), and absorbing the particles that do not move along the path of the aperture. By using a collimation process, this isotropic source of neutrons and gammas from the reactor are able to be shaped into a single beam of fast neutrons.

Collimation Ratio: (2.11)

Where: L=distance between source and image plane D=diameter of the source

Figure 2-4: General image plane for a source showing the image plane and distance (ASTM E803 1996).

There are several components identified in the paper by MacGillivray, some required and some not, which are used to achieve the desired output of a collimator. Each component is comprised of different materials to achieve a different function of the total collimation process, and the type of material depends on the emergent beam in mind. The first step in the process of designing a collimator is to identify the desired output of it so that the appropriate materials and components are formed. Since this projects intention was to make a fast neutron beam using the reactor as a source of neutrons, a similar approach was to that of the upgrade to Beam Port 2.

Figure 2-5 is the 3-D model of the collimator design for upgrade to Beam Port 2. The design included the following components:

(1) Illuminator, typically a moderator that provides a uniformly intense source of

neutrons.

(2) Beam Filter, to remove unwanted radiation. And,

(3) Aperture, defines the pinhole. (McGillivray)

Figure 2-5: Final Collimator design for the Beam Port 2 upgrade in 2012 (Turklogu 2012).

Since high energy neutrons and gammas are produced in the source, an important aspect of this collimator is to remove the unwanted gammas without compromising the neutron flux.

The intention of the filter is to entirely remove the gammas with radiative capture while at the same time allowing neutrons to pass through. This accomplishes two goals; first it reduces the amount of shielding necessary for gamma exposure at the beam exit, and second, it will create a sharper image for radiography. Gamma contributions will act as background noise and can distort the image if not remedied. In neutron imaging, it is not the intensity of the gamma flux itself, but the neutron to gamma ratio (n/γ) that is important in generating a good image.

Generally (MacGillivray n.d.);

This ensures that the image will be generated mainly from neutrons without much gamma contribution.

2.3.1 Illuminator

The Illuminator is used to create a uniformly intense source of neutrons. Since the neutron flux in a reactor can have a wide range of energy, from thermal energies of <1 eV up to several MeVs, it is desirable to shorten this energy range to aid in manipulating the neutrons. As discussed previously, the interaction of a neutron with a compound nucleus is based on the neutron cross-sections. The value of these cross-sections can vary drastically based on the energy of the neutron, and at higher energies there are resonances. In Carbon, for example, the total neutron cross-section can be seen in the figure below. A good illuminator, which is typically a moderator such as graphite, will not only slow down neutrons out of the resonance regions, but also reflect thermal neutrons back into the source. In this way, the neutrons coming from the source are more uniform in intensity.

Figure 2-6: Neutron Cross Sections for Carbon (Lamarsh 2001).

Since the illuminator is a moderator and scatters the neutrons, it needs to be placed as close to the source as possible. If the illuminator were to be placed downstream of the beam, it might scatter the beam and defeat the purpose of the aperture. In addition, the illuminator should have a small clearance so that it does not act as a filter for the neutron flux. Therefore, the illuminator is the first component upstream of the collimator with a small clearance to allow particles to pass.

2.3.2 Beam Filter

Similar in idea to a filter for other applications, the beam filter removes unwanted particles. Unlike a filter used for the oil in your car, however, a beam filter can be designed to remove different particles based on the filter material. Since radiation from the reactor source

17 interacts differently with certain materials, the selection of these materials is imperative to remove the unwanted particles without removing the particles needed for the beam output.

As stated before, the filter choice is based on the ability to transmit the desired neutrons while removing the unwanted gammas. This is achieved based on the various cross-sections of the material. The total cross-section, σtot, is based on the following: (1) σabs, the absorption cross-section, which is typically linearly dependent on wavelength; (2) σel, the coherent Bragg scattering cross-section, which is dependent on neutron wavelength, temperature, orientation and crystal grade ; (3) σinc, the incoherent elastic cross section, which is usually small and independent of wavelength; and (4) σinel, the inelastic cross-section, which is dependent on neutron wavelength and can be significantly reduced with cooling (Mildner 1993). During the upgrade to Beam Port 2, the σtot for the chosen material was desired to be low for thermal neutron energies, and for Beam Port 1, this value is desired to be low for fast neutron energies.

For the collimator in this project, the removal of gammas was desired without removing a significant amount of fast neutrons. Since the majority of neutrons are born as high energy fast neutrons, and the beam ports are facing the fuel plates of the core, there is no need to add a filter to remove these neutrons. During the upgrade to Beam Port 2, it was found that a single filter is not able to remove both gammas and fast neutrons from the beam path. Because of this, two different filters were used for that upgrade.

Comparing the beam outputs between the two port facilities, it was desired to have thermal neutrons shape a beam for Beam Port 2. It used a Bismuth filter for the removal of gammas, and a separate Sapphire filter for the removal of fast neutrons. Using the experience gained from Beam Port 2, only a Bismuth filter is used for gamma removal in Beam Port 1.

2.3.3 Aperture

The Aperture is what defines the pinhole to create the shape of the beam. The pinhole is the path of the beam, so the aperture guides the neutrons along the pinhole while removing the particles that get scattered out of this pinhole. In order to remove the unwanted particles, the aperture needs to be made of different kinds of highly absorbing material. This results in a layering of different materials throughout the aperture. The reason for the different types of material is discussed below in the materials section.

Since this project required a parallel beam, the pinhole is the same size throughout the length of the collimator.

3. Materials

3.1 Collimator Materials

Many of the collimator components had to be replaced in Beam Port 2 due to degradation, which eventually led to the replacement of all of the individual collimator pieces with the exception of the Sapphire crystals. The Polyethylene used for many of the components in the

Beam Port 2 collimator experienced enough degradation that it was causing problems with the shaping of the thermal neutron beam. This led to a redesign of Beam Port 2. Since much of this upgrade was based on the upgrade in 2012, the redesign and use of materials had the same theory behind it, with a few improvements.

3.1.1 Graphite Illuminator

Graphite is typically used as the illuminator, with a thickness ranging from 10-15 cm, and was initially used as an illuminator in the upgrade to Beam Port 2. The measured thermal diffusion length of Carbon at 20o C is 59 cm. (Lamarsh 2001), so the fast neutrons are not in danger of slowing down to thermal energies. However, as the thickness of Graphite increases, there is an increase in the release of gammas due to thermal neutron capture in Carbon. The two isotopes that make up Graphite are 12C and 13C, with the following abundance and cross section data.

Table 3-1: Carbon nuclide data (NNDC n.d.).

Isotope σa (barns) Abundunce (%) Reaction 12 C .00353 98.93 13 C .00137 1.07

The thermal neutron capture cross section of Graphite is quite small, even smaller than

1H with a cross section of .3326 barns. However, with a significant neutron flux, as the length of the Graphite increases, more neutrons are allowed to slow down to thermal energies, and the probability of neutron capture increases, thus adding more gammas to the beam. The energy of the gammas released is not initially important. The more important aspect of the gammas is the addition to the gamma flux which requires more shielding and distorts the imaging in radiography. The Graphite thickness must be utilized so that it can perform its duties as an illuminator without increasing the gamma flux going into the path of the collimator beam.

The upgrade to Beam Port 2 had a Graphite illuminator in the initial design, but was ultimately removed. For that upgrade, a solid block of Graphite machined to fit inside the angled end of the collimator tube was used. During the testing of the beam, it was suspected that the

Graphite significantly reduced the thermal neutron flux at the beam exit. After conducting analysis of the thermal neutron flux at the beam exit using the gold foil activation technique, it was found that the solid block of Graphite reduced the beam by 38.5% when the reactor was at

90% power. This information was taken into consideration for the current upgrade to Beam Port

1. Rather than using a solid block of Graphite, a clearance slightly larger in diameter than that of the aperture is machined concentric to the beam path.

Another consideration for this component is the thermal expansion of Graphite. It has long been recognized that the temperature-dependence of the dimensions of a polycrystalline graphite aggregate must be a function of several factors: (1) the thermal expansion of the crystal lattice, which is markedly anisotropic, and (2) the degree of preferred orientation of the crystals, which depends on the structure of the original coke particles and on the type of manufacturing process (Sutton 1962). Graphite expands anisotropically, along one axis, so for design purposes, the orientation of the piece will determine which axis expands when heated. The heat generated by the OSURR is not necessarily significant enough to worry about thermal expansion of these components; however, they will be exposed to a significant neutron flux, especially the Graphite piece. With that, there is the possibility of Graphite to expand, and it can be manufactured to expand either axially or radially. The original Graphite piece placed inside Beam Port 2 during its initial upgrade was still available, but it was not entirely clear how that piece was manufactured to expand. Because of this, a new graphite piece was made for Beam Port 2 during this project. The new Graphite piece was manufactured to expand axially, and a small gap is designed in the collimator tube to account for this possible expansion.

Both Beam Port 1 and Beam Port 2 have a brand new Graphite piece designed to expand in the axial direction in the event of any type of expansion. The Graphite for Beam Port 1 has the following dimensions: Cylindrical shape with an OD of 5.46”, a length of 5” (12.7 cm), and a concentric clearance with a diameter of 1.5”. This piece is placed inside the aluminum tube that will house the entire collimator, and is placed at the end closest to the core.

Figure 3-1: Graphite, Beam Port 1.

Beam Port 2 was slightly different since the Graphite piece will be placed against the 30o face of the port. It has similar dimensions to the Graphite in Beam Port 1 with the following exceptions. It has a diagonal cut at a 30o angle making the shortest length 2.25”, and the longest length 5.114”, and the OD is slightly smaller at 4.96”. Due to difficulties in manufacturing a way to place an end cap on the aluminum tube at an angle, the Graphite piece is designed for this purpose. This piece is placed inside its own aluminum shell, hence the smaller OD, screwed in place, and then this shell is screwed onto the aluminum collimator tube. This design keeps the

Graphite end exposed to the core open as opposed to being completely surrounded in an aluminum shell like in Beam Port 1.

Figure 3-2: Graphite, Beam Port 2.

Figure 3-3: 3-D image of Graphite with aluminum case for Beam Port 2.

3.1.2 Bismuth and Sapphire Filters

A significant amount of the experience gained from using Bismuth and Sapphire filters for the upgrade to Beam Port 2 was used in choosing and manufacturing the Bismuth for this project. Additionally, a new Bismuth filter was manufactured for Beam Port 2, along with the original Sapphire Crystal filter.

During the previous upgrade to Beam Port 2, the selection of materials based on low σtot for various neutron energies reduced the candidate materials to a select few. These were Quartz

(SiO2), Bismuth, Beryllium, Magnesium Oxide, Silicon and Sapphire (Al2O3). The availability of large single crystals reduced the candidates to Quartz, Sapphire, and Silicon. Sapphire was ultimately chosen because of its increased efficiency over that of quartz and silicon as a thermal neutron filter, even when the latter are cryogenically cooled to 77 K to improve efficiencies

(Nieman 1980). Also, the transmission characteristics of sapphire are not degraded by irradiation for a number of years within a beam port of a reactor or by the use of poorer grade sapphire (Mildner 1998). A single crystal Sapphire is rather costly as well, therefore the sapphire piece is re-used for the current upgrade.

Figure 3-4: Three single-crystal Sapphires (left) used in the 2012 upgrade to Beam Port 2. Those same crystals (right) inside a high density Borated Polyethylene holder (Turklogu 2012).

In order to remove gammas, the most important property of a material is the density.

Sapphire, having a density of roughly 3.98 g/cm3, is not the ideal choice for removing these particles. For normal gamma attenuation, Lead is the best choice with a high density of 11.34 g/cm3. In terms of being used as a filter, Bismuth has a lower neutron-attenuation coefficient than Lead, with nearly identical gamma attenuation (Macgillivray n.d.). Using eq. 2.7, and the mass attenuation coefficient for Bismuth at different energies, a comparison of the gamma attenuation of Bismuth for several thicknesses can be calculated. Figure 3-5 shows this comparison.

Figure 3-5: Gamma Attenuation for different thicknesses of Bismuth.

In practice, the Sapphire crystal has a relatively high scattering cross-section for high energy neutrons, and a much lower neutron absorption cross section, making it ideal for removing fast neutrons by scattering interactions. Bismuth can be used as a neutron filter, however, when selecting the materials during the previous upgrade, it was found that as a neutron filter it has undesirable high values of σinel due to a high density of low frequency phonon states. Therefore it was primarily used as a gamma filter to aid in reducing fast neutron transmission. Additionally, studies have been made for the use of Bismuth as a filter. One such study was conducted in Cairo, Egypt for the transmission of neutrons, ranging in energy from 10-

4 to 10 eV, incident on polycrystalline and imperfect mono-crystals of Bi (Adib 2003). This research concluded that a single crystal of Bi with a thickness of 10.16 cm (4”) will generate the

27 preferred neutron to gamma ratio for this energy range. When comparing this energy range to a larger energy range, it can be seen that the σtot for Bismuth remains relatively steady up to 9 barns (Freund 1983). It is not until it neutrons reach energies around 1 keV where there begins to be significant resonances (NNDC n.d.). The figures below show the total cross-section data for

Bi in several energy ranges.

Figure 3-6: Bismuth total cross-section for 1 - 600 meV neutron energy ranges (Freund 1983).

Figure 3-7: Bismuth total cross-section for 10-3 - 10 eV neutron energy ranges (Adib 2003).

Figure 3-8: Total neutron cross section for 209Bi, from .0125 eV – 5 MeV (NNDC n.d.)

Table 3-2: Comparison of the three filter materials and associated densities and cross sections.

σ σ Material Composition Density (g/cm3) Total scattering absorption (barns) (barns) Sapphire Al2O3 3.98 15.68 .2556 Bismuth Bi 9.78 9.141 .02 Lead Pb 11.34 11.106 .0928

Keeping this in mind, the fabrication of the Bismuth was the next step. The previous upgrade melted Bismuth ingots into a mold and then inserted into a polyethylene holder around the radial edges. This was done to try to save costs. It was found that this process created slight

30 porosity among the surface of the Bismuth mold. In order to avoid this, the Bismuth was purchased as a single piece. Another consideration was the holder for the Bismuth. The previous holder for Bismuth was the aluminum mold that surrounded the radius of the fabricated piece. A new holder was designed to completely encapsulate the Bismuth because there is a slight possibility of 238Pu in the Bismuth. After being exposed to a neutron flux, 209Bi undergoes radiative capture with a neutron. After it captures a neutron it becomes 210Bi, which is an unstable isotope. This isotope will eventually decay into 238Pu over a long decay chain. 238Pu emits a high energy α for 100% of its decays, thus making it a potential radioactive hazard for handling purposes. Because of this possibility, a new aluminum case was designed for the

Bismuth filter for both Beam Ports.

Table 3-3: Bismuth nuclide data (NNDC n.d.).

Isotope σa (barns) Abundunce (%) Reaction 209 Bi .0324 100

Figure 3-9: Poly-crystal Bismuth with molded into an aluminum holder. The piece was slightly machined to remove impurities that surfaced during the melting phase, however, some porous areas remained (Turklogu 2012).

Since the Sapphire crystal was re-used, no new fabrication was required for it, however, a newly designed case was fabricated. The original Sapphire holder consisted of high density

Borated Polyethylene. This holder was Borated to aid in the removal of neutrons after they are scattered by the crystal. Since the new collimator design was desired to be able to last longer without repair, another aluminum case was designed for this filter. The Sapphire crystal was small enough so that a case could be designed to fill it with Borated Cement. The tradeoffs of using Borated Cement and Aluminum as opposed to Borated Polyethylene are discussed later.

Figure 3-10: 3-D image of new Sapphire holder.

The Bismuth filter for the current upgrade must be able to effectively remove the gamma content without removing a significant amount of neutrons through scattering and absorption interactions. Based on the results of Beam Port 2, it was concluded that a similar length Bismuth filter should be the only filter used in Beam Port 1. The dimensions of the Bismuth are a cylinder with an OD of 4.96”, and a thickness of 4”.

Figure 3-11: Bismuth encased in Aluminum Shell.

The final filters being used in Beam Port 2 are both Sapphire and Bismuth, and only a

Bismuth filter for Beam Port 1. Each filter is encapsulated in an aluminum case, and the case for the Sapphire crystal is filled with Borated Cement.

3.1.3 Aperture Pieces

The purpose of the Aperture is to shape the collimated beam through a length of pinholes, while absorbing any particles not moving along the path of those pinholes so they are not scattered back into the beam. Those particles that are required to be absorbed are neutrons, so a strong neutron absorbing material is needed for this. 10B is a very common material used for neutron absorption due to its extremely high σabs at thermal neutron energies, and was used for this upgrade in several configurations. The figure from Knoll shows the σabs of three isotopes typically used for neutrons. A downside in using 10B is that it emits a relatively high energy gamma after absorbing a neutron, and in order to reach this high σabs, the neutrons need to slow down to thermal energies. The below table summarizes these interactions.

Table 3-4: Boron nuclide data (NNDC n.d.).

σ @ ≈.025 eV Reaction Reaction Q-Value Isotope abs Reaction (barns) % (MeV) 10 B 3840 4 2.792 10 B 3840 96 2.31

In 96% of the 10B-neutron reactions, an excited 7Li atom is created. This atom quickly de-excites into stability and releases a 482 keV gamma. These gammas need to be accounted for when

34 developing apertures made of 10B. These absorption interactions also occur in the neutron thermal energy ranges, therefore the neutrons must be moderated, and thus 10B alone is not able to remove these neutrons. Hydrogen is incorporated with 10B to first slow down the neutrons so that they can be removed through absorption.

3 6 10 Figure 3-12: σabs of He, Li, and B for a large range of neutron energies (Knoll 2010).

Both Beam Ports had all of the same aperture pieces, which required an entire replacement to the aperture pieces in Beam Port 2. This was to aid in the modularity and similarity in design of the Beam Ports. Additionally, each collimator was broken up into two sections, an inner collimator, on the reactor side, and an outer collimator, on the reactor bay side. Since the Beam

Ports have the same pieces, the differences of the aperture pieces are between the inner and outer section. These differences are the placement of the aperture pieces between the two collimator

35 sections, as well as the dimensions of the individual pieces. The final layout of the collimators is discussed in the design section.

3.1.3.1 Borated Cement

One of the first departures from the original design of the collimator in Beam Port 2 is the use of Borated Cement as an aperture piece. The original design consisted of high density

Borated Polyethylene. The reason this change was made was because of the degradation of the

Borated Polyethylene. Even though the upgrade was completely in 2012, a relatively short amount of time ago, the Borated Polyethylene is already showing signs of wear from the neutron exposure. This resulted in the change of the beam profile of Beam Port 2 since the aperture holes were beginning to change shape from this wear. Therefore, a more durable design was preferred to prevent replacing the polyethylene pieces every so often.

The downside in using the Borated Cement is the reduction in Hydrogen content, which could result in the reduction of removing the neutrons from the Beam Path. The figure below compares the composition of the two materials. The Boron weight percentage is roughly the same, however, the Hydrogen content is over twice that in the Polyethylene than in the Cement.

In order to justify this replacement, MCNP simulations were run to specifically address this issue. The results concluded that both materials will produce a similar beam output and no neutrons will escape the biological shielding. The MCNP simulations will be discussed later.

Table 3-5: Composition data of Borated Cement vs. Borated Polyethylene (ShieldWerx n.d.).

Composition Data Borated Cement Borated Polyethylene H atomic density (cm-3) 4.76 x 1022 7.48 x 1022 H weight percent (%) 4.73 11.7 B atomic density (cm-3) 4.67 x 1021 2.99 x 1021 B isotope distribution 19.6% 10B and 80.4% 11B 19.6% 10B and 80.4% 11B B weight percent (%) 4.99 5.0 Total density (g/cm3) 1.68 1.07

An Aluminum shell was used to mold the Borated Cement piece and aid in the durability and structure of the piece as well. For the inner collimator, the Borated Cement piece has an OD of 5.46”, a total length of 8”, and a concentric aperture hole of 1.25”. The outer collimator has an OD of 6.46”, a total length of 10”, and the same concentric aperture hole of 1.25”. The below figure shows the assembly of the Borated Cement piece for the inner collimator. The Borated

Cement piece for the outer collimator is assembled in the exact same way. This piece is the first aperture piece after the filter, and the first piece in the aperture pattern for the different layers.

Figure 3-13: Borated Cement piece for Inner Collimator.

3.1.3.2 Metamic® (Borated Aluminum)

Metamic® is the material used in the upgrade to Beam Port 2, and was found to be a very good material for absorbing neutrons as well as gammas. This material was initially developed by the Reynolds Metal Company as a neutron poison in nuclear fuel racks, and is now used as neutron shielding for different types of fuel storage. It is fabricated by Orrvilon, Inc., in Orrville,

OH. which makes this is a good economical material due to the proximity of the fabricator to the

OSURR and there is a pre-existing relationship with this company from previous projects.

Metamic® is a fully-dense, discontinuously-reinforced, metal matrix composite material.

It consists of high-purity Type 6061 aluminum (Al -6061) alloy matrix reinforced with Type 1

ASTM C-750 isotopically-graded boron carbide (B4C). The below figure is a data sheet for different compositions of Metamic®. The composition that was acquired from Orrvilon, Inc. contained a 10% weight ratio of B4C.

Table 3-6: Metamic® composition data from Reynolds Metal Company. This shows the relationships between the 6061 aluminum matrix and the B4C reinforcement for different B4C contents, as well as the resulting densities of the composite materials (Holtec Int. n.d.).

Composite density Wt. %B C Wt. %Al-6061 Vol. %B C 4 4 (g/cm3) 4.68 95.32 5 2.691 9.4 90.60 10 2.682 14.14 85.86 15 2.673 18.92 81.08 20 2.664 23.73 76.27 25 2.655

The Metamic® helps in absorbing the epithermal neutrons that escape the Borated

Cement and Graphite pieces. Since this material will not readily slow down the neutrons, these pieces are placed after the moderator materials. The aluminum will also be able to remove a few gammas since due to its relatively higher density. The shapes of these pieces are ¼” thick disks, with 1.25” concentric aperture holes. The inner collimator has an OD of 5.46”, and the outer collimator has an OD of 6.46”.

Figure 3-14: Metamic® Apertures.

3.1.3.3 Lead

The main purpose of the Lead is to remove the gammas from the 10B(n,α)7Li interaction.

For this reason, it is placed downstream of any Borated materials. Its high density of 11.34 g/cm3 makes it a suitable material for gamma attenuation. These pieces are manufactured from a company called Pure Lead. They are also discs, 1” thick, and each piece also has a 1.25” concentric aperture hole. The inner collimator has an OD of 5.46”, and the outer collimator has an OD of 6.46”.

Figure 3-15: Attenuation coefficients for Lead. (Lamarsh 2001).

Figure 3-16: Lead Apertures.

3.1.4 Aluminum 6061 as Structural Material

Aluminum is a preferred choice to house the collimator pieces due to its relative strength, resistance to corrosion, low neutron absorption cross section, and abundance. It is not the leading candidate in all of these categories individually, but when taken together, it is the better choice. In terms of strength, compared to steel, aluminum is not nearly as strong. When evaluating material strength properties, stress-strain curves are often used, and steel alloys can have strength properties as much as ten times higher than that of aluminum alloys. However,

Steel has two undesirable properties that make Aluminum a better choice, corrosion and neutron activation.

Figure 3-17: Stress-Strain Curves (Oberg 2012).

For this project, corrosion resistance is not a high priority, but to enable a durable long- lasting structure, corrosion is taken into account. The Iron in Steel can cause rust in wet environments as opposed to the Aluminum which is more resistant to corrosion. The environment in the beam ports, as well the reactor bay itself, is dry, so it is unlikely to stimulate significant corrosion. However, it is preferable to have corrosion resistant materials inside the beam port to minimize replacing any components due to corrosion that are exposed to a neutron

42 flux. Most metals will become activated with this kind of neutron flux, so replacing anything will require extra precaution and difficulty.

Aluminum will still become activated when exposed to a neutron flux, but when compared to Steel, the activation is much less. Table 3-8 shows a comparison of several materials with their absorption cross sections at thermal energies. The σabs for aluminum at higher neutron energies can reach values of .6-.7 barns, and in the neutron energy range of 50-

200 KeV, can reach individual resonances of up to 10-15 barns. A typical activation reaction produces the following:

Table 3-7: Aluminum nuclide data (Modulo 1976).

Isotope σa (barns) Abundunce (%) Reaction 27 Al .23 100

27Al will also emit a gamma after becoming 28Al, and 28Al is unstable and undergoes β-decay every time.

Table 3-8: Thermal Neutron σabs for Several Metals (Tsoulfanidis 2013).

Material σabs (barns) Material σabs barns Mg .063 Fe 2.6 Be .01 Mo 2.7 Zr .18 Cr 3.1 Al .23 Cu 3.8 Sn .63 Ni 4.6 Nb 1.10 Ti 5.8

Pure Aluminum is not used for this upgrade, but the Aluminum 6061 alloy is instead. Al-

6061 is a Magnesium-Silicide alloy that is composed of a slight mixture of Mg2Si. The weight percentage for Al-6061 is shown in Table 3-9. The benefits of this is that it is stronger than pure

Aluminum, and magnesium bearing alloys are the only type of alloys that have a lower neutron

σabs than the pure metal. There are stronger Aluminum alloys, but most other alloys contain metals that increase the neutron σabs, and those with Copper increase this by 50%. Al-6061 is a strong Aluminum alloy with the smallest σabs.

Table 3-9: Al-6061 composition with Mg2Si (ASM n.d.). Material Weight Percent Range (%) Aluminum 95.9-98.6 Magnesium .8-1.2 Silicon .4-.8

The Aluminum pieces were used as cases for two filters, the Borated Cement apertures, and the Graphite piece for Beam Port 2, and are shown in Figures 3-3, 3-10, 3-11, and 3-14.

Aluminum was also used to house all of the collimator pieces for each Beam Port. Since each

Beam Port was separated into an inner and outer collimator, each port had two separate aluminum shells to house the collimators. The outer collimator for each Beam Port was the exact same since it was downstream of the Shutter Assembly Box, which had the same dimensions at each port.

3.2 Beam Stop Shielding Materials

The main purpose of the Beam Stop is to shield the high energy neutron and gamma particles exiting the Beam Port from outside the experiment area. Since the aperture pieces have a similar goal in removing the unwanted particles, similar material is used. The basis of these materials and design are the same as what was used for Beam Port 2. However, since this upgrade was for the creation of a fast neutron beam, more external shielding was required. In order to completely remove the neutrons, they must be absorbed, and the most likely energy ranges of neutron absorption are thermal energies. Therefore, more shielding is required to slow down the incoming high energy neutrons so that they can be absorbed.

3.2.1 Borated Polyethylene

The Borated Polyethylene is the main source of neutron removal. Referring back to Fig. in section 3.1.3.1, which shows the material composition of the high density Borated

Polyethylene, this material has high hydrogen content. It is the least dense of the materials being used, so the gamma attenuation will not be as significant, but the high hydrogen content is desired for moderating the fast neutrons. This material is implemented in 1” thick rectangular sheets, and has a height of 30”.

Figure 3-18: Example of Borated Polyethylene sheet drawing.

3.2.2 Metamic®

Metamic® is used for the same purpose as the apertures. It also serves as a transition between the Borated Polyethylene and the Lead. These are also used as rectangular sheets, with a ¼” thickness. These sheets surround the Polyethylene sheets, and have several different configurations. In general, the sheets are 30” high as well.

Figure 3-19: Example of Metamic® sheet drawing.

3.2.3 Lead

The high density of lead makes it a good gamma attenuator. Lead sheets will capture all of the gammas that are produced in the neutron interactions as well as any potential gammas passing through the rest of the shielding material. These are 1” thick rectangular sheets, like the

Poly, and have several configurations as well. These surround the Metamic® sheets and are placed at the outer edges of the Beam Stop. Similar to the other sheet materials, they have a height of 30”.

Figure 3-20: 3-D model of lead sheet.

4. Design

The design for this upgrade used much of the experience and knowledge from the previous upgrade to Beam Port 2 and formed additional goals for improvement. These criteria for the upgrade, apart from shaping the fast neutron beam, are; a more robust structure, similarity between the two Beam Ports, a review process for the design, modularity for each collimator, and the use of existing shielding components.

The robust design is desired so that the components do not need to be replaced as often since they will be exposed to a neutron flux and become activated. The degradation of the polyethylene caused concern since these pieces were in need of being changed out after only five years. Using stronger materials led to a few shielding tradeoffs, such as less moderation of the neutrons, but in the end the stronger construction was of more value.

The similarity helps to reduce cost for fabrication and aids in any possible future changes.

Constructing similar pieces for each collimator allowed a simpler process for reviewing the design as well. The retention of the ideas and drawings for the entire process was emphasized so that the construction of this upgrade could be well documented for personnel in the future. The documentation included drawings that were created for each component of the upgrade. These drawings were reviewed and used to fabricate the pieces. The similarity between the collimators allowed fewer drawings between each Beam Port. For example, the outer collimator for each port was the exact same, so only one set of drawings was needed for both outer collimators. In

49 terms of fabricating, this allowed the manufacturer to produce the same component twice as opposed to creating two separate pieces.

The modularity allows flexibility for future designs. The first upgrade intended to create a vacuum inside by sealing off the collimator portion. This would produce a better beam output since the neutrons would have no chance to collide with the air molecules. In order to accomplish this, the aluminum shell had to be welded together and permanently install the collimator pieces. However, since the pieces were permanently installed, no changes could be made, including replacement of parts in disrepair. This was another tradeoff which resulted in an unsealed aluminum shell, with the ends bolted together as opposed to being welded.

The use of the existing structure resulted in an overall shorter collimator, but better shielding. The existing structure is discussed in section 4.1. The existing structure of the Shutter

Box Assembly caused the collimator to be divided into two sections as opposed to a single collimator passing through the Shutter Box Assembly.

The design of the current upgrade compared to the previous one resulted in a few engineering tradeoffs that needed to be addressed. These were verified using either MCNP simulations or calculations from formulas.

4.1 Existing Structure

4.1.1 Beam Port 2 Facility

The Beam Port 2 Facility was a successful addition to the OSURR. The area external to the beam is able to shield the thermal neutrons and gammas to the outside of the experimentation set-up. In order to do this, the facility had an extensive amount of shielding installed. A separate

Beam Shutter was designed that could be raised and lowered to turn the beam on and off, and a

Beam Stop downstream of the sample area was created to stop the beam itself. The area was also surrounded by cement bricks for extra shielding.

Figure 4-1: Rear View of Beam Port 2 (OSURR n.d.).

Figure 4-2: Overhead view of Beam Port 2 (OSURR n.d.).

4.1.2 Beam Port 1 Current Lay-out

The upgrade to Beam Port 1, as well as the modifications to the Beam Port 2 collimator, are designed with the existing structure and surrounding components in mind. The existing structure includes not only the geometric dimensions of the Beam Ports in the shielding, but a shutter assembly installed into each port, as well as everything in the reactor bay area that is in the immediate vicinity of the port openings. The importance of keeping these things in mind for the design is to utilize the existing components already used to aid in shielding, such as the shutter

53 assembly, as well as ensure a safe working environment where the beam exits. The existing components aid as well as add constraints to the upgrade of the facility.

Figure 4-3: Beam Port 1 at exit of the port (OSURR n.d.).

The Beam Ports themselves extend radially from the fuel plates of the core, parallel to the ground. The circular ports are carved through the biological shielding, and on the reactor side of this shielding, aluminum cylindrical shells, bolted onto the concrete, extend the ports into the reactor pool almost directly next to the fuel plates. This allows the reactor side of the ports to be

54 exposed to a higher flux from the core. As mentioned before, Beam Port 1 extends perpendicular to the fuel plates, while Beam Port 2 extends out from the reactor at a 30 degree angle to Beam

Port 1. The Figures shows an overhead view and 3-D representation of this configuration. The approximate length of Beam Port 1, from measurements taken in 2006, is 80.75”, and the length of Beam Port 2 is 85.25”. The diameter of the ports carved into the cement is 6”.

Figure 4-4: 3-D overhead view of Beam Ports inside the OSURR.

Figure 4-5: 3-D view of Beam Ports inside the OSURR.

These lengths, however, are broken up into two sections due to the installation of a beam shutter assembly in each port. The beam shutter assemblies are boxes, 18” high x 12” wide x

36” long, designed from Lockheed Martin, for additional layers of shielding. The Figures show the Shutter Box when it was removed from the biological shielding. The assemblies are installed in the biological shielding on the bay side, and have their own circular ports that align with the ports carved into the cement. The additional shielding comes in the form of a shutter and radial transition. The shutter is a cylinder, filled with lead shot, and has a 6” diameter hole that can be rotated into an open or shut position. The transition is downstream of the shutter, and increases the diameter of the port from 6” to 7”. The purpose of this is to prevent any radiation leaking along the radial edges of the port when cement plugs, or in these case apertures, of the corresponding diameters are installed.

Figure 4-6: 3-D cut-out of the Shutter Box Assembly in the “Open” (left) and “Shut” (right) position.

Figure 4-7: Removal of the Shutter Box Assembly (OSURR n.d.).

Figure 4-8: Shutter Box Assembly when removed for maintenance (OSURR n.d.)

With the inclusion of the shutter and transition, which requires the collimator to be split into two different sections, each collimator section has different diameters in order to incorporate the original design of the shutter box assembly. For Beam Port 1, the distances on the reactor side of the shutter, from the aluminum extension into the reactor pool to the shutter itself is

44.75”. Beam Port 2 is slightly different since the aluminum extension is not perpendicular to the rest of the port, but has the end attached at a 30 degree angle to the vertical axis. This makes

Beam Port 2 longer overall, but the distance from the shutter to the aluminum extension that is not at an angle is only 42.75” long. Both ports have diameters of 6” on the reactor side of the shutter. The bay side of the shutter is the same for both Beam Ports since this is the location of the shutter box assembly, and these are the same for both ports. The distance from the diameter transition to the opening of the Beam Ports is 16.18”.

The reactor bay at the exit of Beam Port 1 has several structures that add constraints to the external shielding design. The major concern for any additional structure for the Beam Port 1 upgrade is limited space. Beam Port 2 has an external beam shutter as well as a beam stop directly in front. Between these two structures is the area set up for experimentation, and surrounding all of this is additional shielding in the form of concrete blocks. This limits the space to the left of Beam Port 1. Another space limitation is the sub-critical assembly storage area. This is against the wall across from Beam Port 1. The Figures in these sections show a clearer representation of these space limitations. To prevent tripping hazards for egress through the area, this limits the length of the rails that can be used. In order to use the length of rail that we need, the sub-critical assembly was removed.

4.2 MCNP Simulations

MCNP was used in verifying different shapes and materials that were to be used in not only shaping the fast neutron beam, but also ensuring adequate shielding was being used to stop the beam outside of the experiment area. Since the various simulations were not the work of this author, the exact results will not be discussed, but instead an overview of the set-up and conclusions. However, the final collimator design was created in MCNP in order to give a representation of the model used in the MCNP simulations.

Figure 4-9: MCNP model of the collimator

The MCNP simulations had several designs each for the Collimator and for the Beam

Stop. For the Collimator, these included designs for different size apertures, Borated Cement vs.

Polyethylene, and the use of the shutter in the shutter box assembly. The different designs for the Beam Stop were the use of Borated Cement vice Polyethylene again, and different sizes needed for shielding.

In addition to the different designs, both gammas and neutrons at different energies were simulated. The neutrons were used to find an effective aperture size and adequate Beam Stop shielding. The gamma particles were used to verify the filter thickness, as well as shielding for the Beam Stop. For neutrons, the energy ranged from thermal energies up to 3 MeV, and for gammas, the energy went as high as 6 MeV.

4.2.1 Collimator MCNP Simulations

The original upgrade to Beam Port 2 was able to create a thermal neutron beam using an aperture size of 30 mm (≈1.18”). For this current upgrade, we wanted a larger diameter beam, and ran MCNP simulations for larger diameter apertures. These were in the range of 30mm up to 1.5” (38.1mm). In the end, the 1.5” diameter aperture had too much scattering at the beam exit, so an aperture diameter of 1.25” (31.75mm) was established. This gave a slightly larger diameter beam, and limiting the scattering neutrons moving non-parallel to the beam.

Another feature that was compromised in the original upgrade to Beam Port 2 was the use of the Shutter Box Assembly. Since the collimator tube originally extended through the

Shutter Box Assembly, the Shutter, which aids in shielding when the port is not in operation, was unable to be used. The reason for this extension was to increase the length of the aperture in order to increase the collimation ratio. At the same time, this tradeoff removes designed

61 shielding, which may be needed. One MCNP variation determined the effectiveness of the

Shutter by simulating a collimator with and without it. This concluded that separating the collimator into two pieces will not drastically affect the beam as previously feared, while still maintaining the use of the Shutter.

The last design parameter tested for the collimator is the use of Borated Cement. As mentioned previously, the use of Borated Cement has pros and cons when replacing the Borated

Polyethylene. The tradeoff of using the cement was determined by MCNP simulations.

4.2.2 Beam Stop MCNP Simulations

Similar to the Collimator, the use of Borated Cement was attempted for the Beam Stop.

Since the Hydrogen content of the cement is significantly reduced, the attenuation of neutrons was also reduced. The simulation verified that the cement is not an adequate substitute for shielding the fast neutrons exiting the collimator.

The first design of the collimator was much smaller than the current one in order to try to utilize the space in the bay immediately outside the ports. Originally, a smaller area was going to be used for the Beam Stop due to the sub-critical assembly storage unit. After several conservative MCNP simulations, this area would not be adequate to build the correct size Beam

Stop to stop the fast neutrons and gamma energies. Because of this, the design of the Beam Stop and rail system was much larger than anticipated, and required to be built after the sub-critical assembly was removed.

4.3 Design Calculations

4.3.1 Collimator Calculations

The collimation ratio, eq. 2.11, can be used to provide several key equations in the design of a collimator. The calculation of this ratio for Beam Port 1 is based on physical measurements.

A more accurate technique is based upon analysis of a neutron radiographic image using ASTM

E803. Since the facility has not been set-up for radiography, imaging cannot be analyzed to determine the L/D ratio with this method. The physical measurements used the length of the collimator starting at the beginning of the apertures, and assumed the image plane would extend

6” from the exit of the Beam Port.

The first of these equations uses the collimation ratio to calculate the “neutron flux at the image plane” (MacGillivray).

Neutron flux at image plane ( υ ): (4.1) i ( )

Where: υa = neutron flux at aperture The flux at the aperture is first attenuated through the Bismuth filter. The attenuation can be found by using equations 2.2 and 2.3. For Bismuth, the cross-sections for 209Bi are used since this isotope is 100% abundant in Bismuth. The atomic density of Bismuth, N209-Bi, is found from the following:

Atomic density (N209-Bi): (4.2)

3 Where: ρ209-Bi = 9.78 g/cm

Av = 6.022214 e23 atoms/mol

M209-Bi = 209 g/mol

Using data from the OSURR of the neutron flux at Beam Port 1, Figure 2-1, the total neutron flux is 4.2 e12 n/cm2/s. Since 46% are fast neutrons, the total fast neutron flux is 1.932 e12 n/cm2/s. Since Bismuth has many large resonances, for a thickness of 4” (10.16 cm), the transmittance value is plotted using the ENDF/B-VII.1 library from the National Nuclear Data

Center (NNDC n.d.), Figure 3-8.

Table 4-1: Collimation ratio

Outer Total Length Aperture Inner Collimator Length of Collimation Collimator to Image Dia. (D) Aperture Length Shutter Ratio Length Plane (L) 1.25” 32.75” 15” 16” 63.75” 51

Figure 4-10: Neutron transmittance through 4” of Bismuth from neutron energies of .0125 eV – 10 MeV.

Figure 4-11: Neutron transmittance showing the energy range corresponding to 1/v cross section range (.0125 eV – 1 keV).

In order to simplify the amount of flux transmitted through the Bismuth, an assumption was made about the energy of the neutrons. The assumption is that the Graphite illuminator will reduce the neutrons into the 1/v energy range so that the flux avoids the resonances. Once the flux of the transmitted neutrons is found, equation 4.1 can be used to find the estimated neutron flux at the image plane.

Figure 4-12: Final neutron flux through Bismuth.

Figure 4-13: Collimated fast neutron flux.

4.3.2 Shielding Calculations

4.3.2.1 Fast Neutrons

The first step used in shielding was to find out the desired maximum dose rate corresponding to the final neutron fast flux. In order to maintain the federally regulated limits, a maximum fast-neutron dose rate of .01 mSv/hr (1 mrem/hr) was chosen, and using the data from

Table 9.5 from Cember, the fast neutron flux corresponding to this dose rate is 3.7 n/cm-2-s-1.

With the final fast neutron flux known, and using the collimated fast neutron flux, the number of Half-Value Thicknesses (HVL) can be calculated. This is the number of thicknesses that will reduce the initial flux by half. This is related to the attenuation equations in section 2 by equating the neutron or gamma transmittance to ½.

HVL: (4.3)

Number of HVLs (n): (4.4)

ΣR for the shielding components is not just a single isotope, but a mixture of several materials. In order to find ΣR for the shielding material, the following equation was used.

Removal cross-section (ΣR): ∑ ( ⁄ ) (4.5)

3 Where: wi = partial density (g/cm )

ΣR/ρ = mass removal cross-section ρ = material density

Table 4-2 through 4-4 shows the calculated ΣR for 5% Borated Polyethylene, Lead, and

Metamic®. The weight fractions for each element were used from the data sheet for the given

67 material. Borated Polyethylene weight fractions were found from ShieldWerx, and Metamic® weight fractions were found from Holtec International.

3 Table 4-2: ΣR for Borated Polyethylene (ρ = 1.07 g/cm ).

2 3 -1 Element ΣR/ρ (cm /g) Weight Fraction wi (g/cm ) ΣR (cm ) H .598 .117 .1252 .0749 C .0502 .833 .8913 .0447 B .0575 .05 .0535 .0031 Total .1227

® 3 Table 4-3: ΣR for Metamic (90.6% Al-6061, 9.4% B4C); ρ = 2.682 g/cm ). Weight Weight Σ /ρ Total Weight w Σ (cm- Element R Fraction in Fraction in i R (cm2/g) Fraction (g/cm3) 1) 6061 Al B4C B .0575 0 .772 .0726 .1947 .0112 C .0502 0 .215 .0202 .0542 .0027 8.046e- 1.838e- Ca .0243 0 .003 .0003 4 5 Al .0293 .976 0 .8843 2.3717 .0695 Mg .0333 .012 0 .0109 .0292 .001 Si .0295 .008 .005 .0077 .0207 .006 Fe .0214 .0015 .005 .0018 .0048 1.05e-4 Zn .0183 .0025 0 .0023 .0062 1.11e-4 Total .0852

3 Table 4-4: ΣR for Lead (ρ = 11.34 g/cm ).

2 3 -1 Element ΣR/ρ (cm /g) Weight Fraction wi (g/cm ) ΣR (cm ) Pb .0104 1 1 .1179

Table 4-5: Fast neutron HVL for shielding components.

-1 Shielding Component ΣR (cm ) HVL (cm) Borated Polyethylene .1227 5.649 Metamic® .0852 8.136 Lead .1179 5.879

In order to find the shielding thickness (n*HVL), the build-up factor for neutrons was added, as well as the inverse square law as distance increases from the source. The equation for the fast-neutron flux passing through a material of thickness n*HVL becomes (Cember):

(4.6)

Assuming that the majority of the neutrons will be attenuated in the Borated

Polyethylene, only the HVL for this shielding material will be used. Second, the collimated flux is a conservative estimate of 2e7 n/cm2s-1. According to Cember, a hydrogenous shield at least

20 cm thick will have a dose-buildup factor of approximately 5. Solving equation 4.6 for n will give the number of HVL, which can then be found to find the amount of shielding thickness required. MATLAB was used to solve equation 4.6 for n.

Table 4-6: Shielding thickness for fast neutrons.

Shielding Component n HVL (cm) Thickness (cm) Borated Polyethylene ≈ 10 5.649 56.49 (22.24”)

4.3.2.2 Gammas

The biggest concern for gammas in this shielding is the gammas resulting from the absorption reactions. Assuming all of these gamma rays are absorbed by 10B, and, as previously stated, 10B emits a 472 keV gamma in 96% of its absorptions. Therefore 96% of the collimated neutron flux is used as the gamma flux, and assuming all of it is absorbed in 23” (58.42 cm) of

Borated Polyethylene. For a conservative estimate, the total flux will be calculated at the end of the 23” of Borated Polyethylene, and equations 2.9 and 2.10 will be used to find the exposure after traveling through 4” (10.16 cm) of Lead. The exposure is expressed in Roentgen (R), and 1

R is equivalent to .877 rad in air. The dose-buildup factor for a 472 keV gamma ray is again found in Cember. For dose, the Radiation Weighting Factor (wR) is used to convert rads to rem, and for gammas, this value is 1.

Table 4-7: Gamma dose due to neutron capture reactions.

(μ /ρ)air μPb @ a Exposure, Exposure, γ-flux Relaxation @ 472 Dose 472 keV B(μx) unshielded shielded (cm-2s-1) Lengths, μx keV (mrem/hr) (cm-1) (mR/hr) (mR/hr) (cm2/g) 16.706 1.92e7 1.6443 .0297 ≈1.2 17737.2 1.182 e-3 1.35 e-3 (≈17)

4.4 Collimator Final Designs

4.4.1 Design Process

A review and retention process was used for this work. Drawings were created for each part, and many records were retained such as e-mails and meeting notes. On top of that, an

70 online drive, which every collaborator had access to, was used to upload all of the relevant files for the project.

Each part required a drawing which was used for review. The simpler parts were just a single drawing, while the assemblies, such as the aluminum shells, required several drawings.

Once the drawings were reviewed they were used to fabricate the piece from the fabrication company.

Nearly every single document, from the drawings above, to digital correspondence, was retained. This was to aid in any future design modifications. The majority of these files were also centrally located on a secure drive online. The intention was to allow anyone looking into this project in the future be able to fully and easily understand the choices made for the project.

Figure 4-14: Example of drawing assembly.

Figure 4-15: Example of individual drawing used for above assembly.

4.4.2 Beam Port 1

4.4.2.1 Inner Collimator

The final design of the Inner Collimator for Beam Port 1 consists of the illuminator,

Bismuth filter, and the aperture layers, encased in an aluminum shell. The total length of the collimator, including the aluminum shell, is 42.25”, with an OD of 5.92”. The length of the collimator assembly inside the aluminum shell is 40.75”. The OD for the collimator pieces were

5.46” in order to fit inside the aluminum shell.

Figure 4-16: Inner Collimator, Beam Port 1.

Table 4-8: Summary of material thicknesses, Inner Collimator, Beam Port 1.

Material Total Length (Including Aluminum Shell) No. of pieces Graphite 5” 1 Bismuth 4.125” 1 Borated Cement 8” 3 Metamic® .25” 4 Lead 1” 6

Figure 4-17: Inner Collimator Aluminum Shell, Beam Port 1. The last two pictures show the insertion of the collimator pieces into the shell.

4.4.2.2 Outer Collimator

The Outer Collimator is the same for both Beam Ports. It only consists of aperture pieces. The total length is 16”, with an OD of 6.92”. The length of the aperture assembly inside is 14.5”, with an OD of 6.46” for fit inside the aluminum shell.

Figure 4-18: Outer Collimator.

Table 4-9: Summary of material thicknesses, Outer Collimator.

Material Total Length (Including Aluminum Shell) No. of pieces Borated Cement 10” 1 Metamic® .25” 2 Lead 1” 4

Figure 4-19: Inner Collimator Aluminum Shell, the last two pictures show the insertion of the collimator pieces into the shell.

4.4.3 Beam Port 2

4.4.3.1 Inner Collimator

The Inner Collimator for Beam Port 2 consists of a Graphite illuminator cut at a 30o angle, a Bismuth and Sapphire filter, and the aperture layers encased in an aluminum shell. Since the reactor end of Beam Port 2 was at a 30o angle, the dimensions have a short and long end.

The OD was 5.92” for the entire assembly, including the aluminum shell, and the lengths for

Beam Port 2 is 42.125” on the short end, and 45.54” on the long end. The collimator itself has an OD of 5.46”, with lengths of 40.625” and 43.78”.

Figure 4-20: Inner Collimator, Beam Port 2.

Table 4-10: Summary of material thicknesses, Inner Collimator, Beam Port 2.

Material Total Length (Including Aluminum Shell) No. of pieces Graphite 2.25” (short)/ 5.114” (long) 1 Sapphire 5.125” 1 Bismuth 4.125” 1 Borated Cement 8” 3 Metamic® .25” 4 Lead 1” 4

Figure 4-21: Inner Collimator Aluminum Shell, Beam Port 2. The last two pictures show the insertion of the collimator pieces into the shell.

4.4.4 Aluminum Structural Considerations

Additional considerations were made for the structure of the collimator pieces. This included strength and weight calculations, as well as configuring the design to be screwed together.

The ability to remove the collimators from the ports was imperative in order to accomplish the design features set out in this project. The facility currently uses cement plugs as shielding inside the port when not in use, and has the ability to remove them with a special tool.

With the dimensions of this tool in mind, a cap was specifically designed with a removal hole to allow the use of this tool.

Figure 4-22: Shield plug handling tool end (OSURR n.d.).

Figure 4-23: 3-D model of Aluminum shell cap design for removing the collimator.

Since there is a 12” long Aluminum extension into the reactor pool at each end of the ports, stress calculations were made to ensure no significant amount of force was applied to this extension. The reason for this was the concern that inserting a heavy Graphite and Bismuth piece into the end of the collimator could cause enough bend and shear off the Aluminum, causing the reactor pool water to flood into the reactor bay. This is why the original upgrade had a spacer between the first collimator component and the end of the aluminum shell. To counteract this possibility, stress calculations were made assuming the aluminum shell for the collimator must support the entire load of the Graphite and Bismuth pieces without significant bending. Using the Machinery‟s Handbook, Ch. 12 discusses the various types of cases for stresses and deflections in Beams. The basic equations for the volume, mass, and force were as follows:

Volume (V):

Mass (m): Force (F): F = mg Where: R = Radius L = Length ρ = Density g = acceleration due to gravity

Figure 4-24: Stress and beam load case for Aluminum shell extending into the reactor pool (Oberg 2012).

This resulted in the following sets of equations:

( ) Moment of Inertia (I):

Section modulus of cross-section (Z):

Stress at critical point, support (s):

Max Deflection (De):

Where: OD = Outside Diameter ID = Inside Diameter E = Modulus of Elasticity (10000 kpsi for Al-6061). Additionally, the Aluminum shell was design to be removed, which might cause twisting and additional torsion. For that, the following equations were used:

Twisting Moment (T):

( )

Torsion/Shear Stress (τ): ( )

Angular Deflection (θ):

Where: G = Modulus of rigidity (3900 kpsi for Al-6061) These calculations were made for an Aluminum shell with thicknesses of ¼” and ⅛”.

The Yield Stress for Al-6061 is 40,000 psi, and the Ultimate Stress is 45,000 psi. The results of the calculated stresses and deflections are in Table 4-11 below.

Table 4-11: Max stress and deflection calculations.

⅛” thick ¼” thick

Max Stress (psi) 259.77 104.12

Max Deflection (in) 4.21e-4 1.82e-4

Shear Stress (psi) 400 173

Angular deflection (degrees) 9.18e-3 9.18e-3

4.5 External Shielding Design

The external shielding is comprised of the Borated Polyethylene, Metamic®, and Lead sheets in a block configuration, called the Beam Stop, mounted on a mechanical system that allows the it to move horizontally and vertically. The mobility of the shielding ultimately aids in maintaining access to the ports while also shielding the high energy particles from outside of the

80 experimentation area. When the Beam Port 1 facility is not in use for experimentation, the Beam

Stop rests directly against the concrete biological shielding at the opening of the port. Horizontal motion allows the Beam Stop to be moved away from this wall to allow an experimentation area to be set-up.

The normal orientation of the Beam Stop is for the front face to be perpendicular and centered with the incoming fast neutron beam. While this shields all of the radiation, it limits access to the Beam Port. The vertical movement allows the top of the Beam Stop to become level with the bottom of the port hole. This allows access to the Beam Port without completely removing the shielding to allow any possible future changes to the collimator, or to remove it if need be.

4.5.1 Beam Stop

The Beam Stop is oriented like a box with materials used in the form of sheets. The majority of these sheets are the 1” thick Borated Polyethylene, of which there are a total of 29. 3

¼” thick Metamic® sheets surround these Borated Poly sheets, two on the side and one in the back, and 3 1” thick Lead sheets are placed against the Metamic®. The dimensions of the Beam

Stop are close to a 30” by 30” by 30” box.

Figure 4-25: 3-D Model of Beam Stop.

The first 3 sheets of the Borated Polyethylene have 4” diameter holes drilled into them.

These holes are aligned with the fast neutron beam, and the purpose is to capture the back scattered radiation after it comes into contact with the shielding. At the back end of the Beam

Stop are few lead square blocks inserted into the last 3 Borated Polyethylene sheets. These are

1” thick, and are 8” by 8”. They are in-line with the fast neutron beam as well. During the

MCNP modeling, the Borated Polyethylene was not able to attenuate gammas at several energies that are coming from the reactor via the beam. Therefore, extra lead shielding was placed at

82 back of the Beam Stop to capture these. The Lead was made into the square disks at the back end instead of completely replacing the Borated Polyethylene sheets to; a.) minimize the weight of the Beam Stop, and b.) maximize the use of Borated Polyethylene shielding that can capture the scattered neutrons.

Figure 4-26: 3-D model of Beam Stop cut-out showing Lead inserted in the back.

A significant amount of shielding was needed for fast neutrons, which lead to a much bigger Beam Stop than anticipated. The total weight of this Beam Stop was estimated using the volume, number of sheets, and the density of the sheet material. These values are compiled in

Table 4-6. The total weight includes the additional square Lead disks at the back. Lead contributes a significant amount of the weight, which caused changes to some of the initial designs. The MCNP simulations allowed the determination of an adequate thickness of Lead to minimize the weight of the Beam Stop.

An additional hole is created for the use of the lifting mechanism, and is discussed in the next section.

Table 4-12: Beam Stop estimated weight (weight values are rounded). Material ρ (g/cm3) ρ (lb/in3) No. of Sheets Weight (lbs) Borated 1.07 .0387 29 930 Polyethylene Metamic® 2.67 .096 3 64 Lead 11.34 .41 3 1294 Total 2288

4.5.2 Rail/Lift System

The mechanical system was designed in order to move the Beam Stop horizontally and vertically. Since the Beam Stop for this upgrade was also significantly heavier than the Beam

Stop for Beam Port 2, the mobile system was much different as well.

4.5.2.1 PBC Linear Components

The horizontal motion is accomplished via a roller pillow block linear bearing rail system from PBC Linear. The roller pillow blocks are ideal for heavy loads, and a 1-1/2” diameter shafting has the preferred load rating for the weight of this Beam Stop. Additional shafting and bearing components from PBC Linear were used as a guide system to raise and lower the Beam

Stop.

The rails were also separated into two 3‟ sections. The first section, closest to the beam, is permanently installed. The back section will be removed when the Beam Port facility is not in use so that the rails do not act as a tripping hazard.

Figure 4-27: Double roller pillow block data from PBC Linear (PBC Linear n.d.).

Figure 4-28: 3-D model of PBC linear components with bottom frame.

Figure 4-29: 3-D model showing the removable section of railing

4.5.2.2 Joyce-Dayton Components

The vertical motion required a component rated to lift at least 1 ton, but also maintain the lifted position in the event of any kind of electrical failure. The lift also needed to have relatively small dimensions due to the space limitations. The 2012 upgrade used a lifting column, which is hydraulic, for the Beam Shutter. This lifting column had weight limitations causing the Beam Shutter to be undersized, which in turn required it to be surrounded by permanent shielding in the form of HDPE, BPE, and concrete blocks. The permanent shielding required a larger area in order to be effective, and therefore a similar lifting column could not be used for the current upgrade. For these reasons, a mechanical lift was chosen in the form of a machine screw jack from Joyce-Dayton.

Figure 4-30: Beam Port 2 (Left) Beam Shutter without Lead components, (Right) 3-D model of Beam Shutter with Lead components (Turkoglu 2012).

A 3-ton machine screw jack, keyed for a traveling nut, allows for a 15” rise of the load.

A 120 VAC motor is attached directly to one side of the gear box, and a limit switch is attached to the other.

A keyed for traveling nut jack (sometimes referred to as a rotating screw jack) features a lifting screw keyed to the worm gear as a single unit, forcing the lifting screws to rotate, but not translate. A flanged traveling nut, attached to the load, is driven by the rotation of the lifting screw (PBC Linear n.d.).

Figure 4-31: Typical machine screw jack with a lifting screw (Joyce/Dayton n.d.).

Figure 4-32: Machine screw jack with a traveling nut (Joyce/Dayton 2012).

The rotating screw jack remains vertical. Therefore, a hole was cut out of the Back Stop to allow the screw to pass through. The length of the screw is low enough to not come into direct contact of the incoming fast neutron beam, but high enough to ensure there is enough travel for the Beam Stop to completely raise and lower. The following Table and Figures show information for this specific type of lifting screw.

Table 4-13: Lifting Screw Data (Dayton/Joyce n.d.).

Screw Diameter (in) 1 Max Load (tons) 3 Gear Ratio 25:1 Rise (in.) 15 Linear (Travel) Speed (in/min) 17.5

Table 4-14: Lifting Screw Performance/Safety Details (Dayton/Joyce n.d.).

Input Speed (rpm) 1750 Screw Torque, Raising (in-lbs) 243.642 Input Power, Raising (HP) .608 Screw Max Shear Stress (lb/in2) 4,887 Screw Max Tensile Stress (lb/in2) 8,149

Figure 4-33: 3-D model of Joyce-Dayton Components: (1) Motor, (2) Traveling nut lift screw, and (3) Limit switch

4.5.2.3 Fabricated Components

The remainder of the components were either fabricated, or bought in pieces to be installed individually. The fabricated portion comprised of a bottom frame and middle frame.

Both are made from stainless steel for both structural support and to minimize corrosion. The other parts were for the actual movement of the system. A geared rack and pinion system was designed, similar to the system for moving the shielding of the thermal column, to aid in moving the Beam Stop horizontally.

Figure 4-34: 3-D model of bottom frame (left) and middle frame (right).

Figure 4-35: 3-D Model of all components without (left) and with (right) the Beam Stop.

4.6 Future Work

Apart from these components, several other pieces of equipment are under consideration to complete this upgrade.

One important future addition is the replacement of the Shutter Box Assembly. As discussed previously, the Shutter Box Assembly is operated by the use of a hand wheel, located between the two Beam Port facilities. This can be seen in Figures 4-7 and 4-8. In order to have

91 adequate room for safe operation, the hand wheel must be removed. This will be accomplished by fabricating a completely new, automated shutter system, based on the original Lockheed-

Martin drawing. This will allow a controller to be placed a safe distance from the operation of the moving equipment.

Several other additional upgrades include: permanent radiation detectors that will surround the Beam Port exit and critical exposure areas; portable experimentation area that is simple to set-up and use; and a fast neutron imaging camera for radiography. These will be beneficial to the safety and ease of operation for any future experiments.

5. Conclusions

The progress of this work includes the design of Collimators for both Beam Port facilities, and a Beam Stop for Beam Port 1. The proposed upgrade will not only allow the

OSURR to have a fast neutron beam external to the biological shielding of the reactor, but will also maintain the neutron and gamma dose rates well within the radiological limits required by the NRC.

This does not include the construction and characterization of the fast neutron beam, which will be needed to verify the design. Several different iterations were discussed before these designs were finalized. The upgrade to Beam Port 2 in 2012 was an excellent basis for the approach to this upgrade.

This upgrade improved upon the design of the Beam Port 2 facility by using the same theoretical concepts. Similar components were used for the collimator, such as an illuminator, filter, and aperture, but using different materials and configurations. The new materials aided in a collimator that will be able to last longer. Additionally, alterations in the design changed some goals of the facilities. A few examples are the removal of a sealed collimator for the original upgrade in favor of a collimator shell that can be easily unassembled to change the pieces inside, and dividing the collimator into two pieces to make use of the Shutter Box Assembly.

After characterizing the thermal beam in Beam Port 2, it was found that a solid Graphite illuminator was more of a hindrance to the collimated beam. It cannot be determined how

93 helpful a new Graphite illuminator, this time with a clearance hole, will be until experimental testing similar to the previous upgrade is accomplished. It is also difficult to determine if the amount of shielding needed is adequate until final testing is done as well.

The design process was rather thorough so that future work, after the upgrade is complete, can be accomplished easier. This included things such as record retention, an on-line file drive, and drawings for part fabrication. The drawings were used to fabricate the parts, but also as a technical tool to understand the design.

The original upgrade to Beam Port 2 created a thermal neutron beam, outside of the biological shielding of the OSURR, allowing additional research capabilities such as thermal neutron imaging and neutron depth profiling (NDP). Since this is a thermal neutron beam, it has different interaction cross sections, for many materials, than higher energy fast neutrons. Once the upgrade to Beam Port 1 is complete, the OSURR will have the ability to simultaneously experiment with a fast and thermal neutron beam.

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Appendix A: Drawings Index

Table A-1: Drawing Index

Drawing No. Project Title

604 BP 1/BP 2 Bismuth Plug Assembly

604-A BP 1/BP 2 Bismuth

604-B BP 1/BP 2 Bismuth Plug End Cap

604-C BP 1/BP 2 Bismuth Plug Outer Shell

605 BP 1/BP 2 Cement Plug Assembly, Inner Collimator

605-A BP 1/BP 2 Cement Plug End Cap, Inner Collimator

605-B BP 1/BP 2 Cement Plug Outer Shell, Inner Collimator

605-C BP 1/BP 2 Cement Plug Inner Shell, Inner Collimator

606 BP 1/BP 2 Lead Aperture, Inner Collimator

607 BP 1/BP 2 Metamic Aperture, Inner Collimator

608 BP 1 Graphite, Beam Port 1

609 BP 1 6" OD Aluminum Shell Assembly, Beam Port 1

609-A BP 1 6" OD Aluminum Shell Front Cap, Beam Port 1

609-B BP 1 6" OD Aluminum Shell Tube, Beam Port 1

609-C BP 1 6" OD Aluminum Shell End Cap, Beam Port 1

610 BP 1/BP 2 Cement Plug Assembly, Outer Collimator

Continued

Table A-1 Continued

610-A BP 1/BP 2 Cement Plug End Cap, Outer Collimator

610-B BP 1/BP 2 Cement Plug Outer Shell, Outer Collimator

610-C BP 1/BP 2 Cement Plug Inner Shell, Outer Collimator

611 BP 1/BP 2 Lead Aperture, Outer Collimator

611-A BP 1/BP 2 Lead Aperture, Outer Collimator (7/8")

612 BP 1/BP 2 Metamic Aperture, Outer Collimator

613 BP 1/BP 2 7" OD Aluminum Shell Assembly

613-A BP 1/BP 2 7" OD Aluminum Shell Front Cap

613-B BP 1/BP 2 7" OD Aluminum Shell Tube

613-C BP 1/BP 2 7" OD Aluminum Shell End Cap, Beam Port 1

613-D BP 2 7" OD Aluminum Shell End Cap, Beam Port 2

614 BP 2 Sapphire Holder Assembly, Beam Port 2

614-A BP 2 Sapphire, Beam Port 2

614-B BP 2 Sapphire Holder Inner Cap, Beam Port 2

614-C BP 2 Sapphire Holder Inner Shell, Beam Port 2

614-D BP 2 Sapphire Holder Outer Shell, Beam Port 2

614-E BP 2 Sapphire Holder Outer Cap, Beam Port 2

615 BP 2 Graphite Plug Assembly, Beam Port 2

615-A BP 2 Graphite, Beam Port 2

615-B BP 2 Graphite Plug End Cap, Beam Port 2

615-C BP 2 Graphite Plug Outer Shell, Beam Port 2

616 BP 2 6" OD Aluminum Shell Assembly, Beam Port 2

616-A BP 2 6" OD Aluminum Shell Tube, Beam Port 2

Continued

100

Table A-1 continued

616-B BP 2 6" OD Aluminum Shell End Cap, Beam Port 2

617 BP 1/BP 2 Aluminum Shim, Inner Collimator

618 BP 1/BP 2 Aluminum Shim, Outer Collimator

619 BP 1 1.5" ID Metamic Aperture, Inner Collimator

621 BP1 Beam Stop Rail System, Middle Support Assembly

621-A BP1 Beam Stop Rail System, Middle Support Plate

621-B BP1 Beam Stop Rail System, Middle Support, Angle Iron Support

621-C BP1 Beam Stop Rail System, Middle Support, Angle Iron Support Motor Side

621-D BP1 Beam Stop Rail System, Middle Support, Angle Iron Side Alignment

621-E BP1 Beam Stop Rail System, Middle Support, Angle Iron Front/Back Alignment

622 BP1 Beam Stop Rail System, Bottom Support Assembly

622-A BP1 Beam Stop Rail System, Bottom Support, Center Square Tube

622-B BP1 Beam Stop Rail System, Bottom Support, Front/Back Square Tube

622-C BP1 Beam Stop Rail System, Bottom Support, Angle Iron Left Side

622-D BP1 Beam Stop Rail System, Bottom Support, Angle Iron Right Side

622-E BP1 Beam Stop Rail System, Bottom Support, Center U Channel

622-F BP1 Beam Stop Rail System, Bottom Support, Lower Bearing Mount - Right

622-G BP1 Beam Stop Rail System, Bottom Support, Lower Bearing Mount - Left

623-A BP1 Beam Stop Rail System, Bottom Support, Lower Bearing Alignment

623-B BP1 Beam Stop Rail System, Bottom Support, Upper Bearing Alignment

624 BP1 Beam Stop Rail System, Bottom Support, Lift Spacer

626-A BP1 Beam Stop, Polyethylene Sheet

626-B BP1 Beam Stop, Polyethylene Sheet - Front

Continued

101

Table A-1 continued

626-C BP1 Beam Stop, Polyethylene Sheet - Back

626-D BP1 Beam Stop, Polyethylene Sheet - Middle

627-A BP1 Beam Stop, Metamic Sheet - Side

627-B BP1 Beam Stop, Metamic Sheet - End

628-A BP1 Beam Stop, Lead Sheet - Side

628-B BP1 Beam Stop, Lead Sheet - End

102

Appendix B: MATLAB code for Table 4-6

Neutron shielding solution for n

B=5; S=2e7; flux=3.7; HVL=5.649; A=(B*S)/(flux*4*pi*HVL^2); syms n; eqn=n^2*2^n==A; vpasolve(eqn,n)

103

Appendix C: MCNP code for Figure 4-9 c CELLS 1 236 -1.7 -2 32 7 -14 $Graphite 2 457 -2.64 -2 5 14 -15 5 $Metamic #1 3 208 -2.699 -2 15 -28 $Bismuth Shell Back End Cap 4 208 -2.699 -2 29 -30 $Bismuth Shell Front End Cap 5 208 -2.699 -2 27 28 -29 $Bismuth Shell Outer Tube 6 310 -9.747 -27 28 -29 $Bismuth 7 208 -2.699 -2 5 30 -33 $Cement #1 Back End Cap 8 208 -2.699 -2 5 34 -35 $Cement #1 Front End Cap 9 208 -2.699 -2 31 33 -34 $Cement #1 Outer Shell 10 208 -2.699 -32 5 33 -34 $Cement #1 Inner Shell 11 228 -3.35 -31 32 33 -34 $Cement #1 12 457 -2.64 -2 5 35 -16 $Metamic #2 13 252 -11.35 -2 5 16 -17 $Lead #1 14 252 -11.35 -2 5 17 -18 $Lead #2 15 457 -2.64 -2 5 18 -19 $Metamic #3 16 208 -2.699 -2 5 19 -36 $Cement #2 Back End Cap 17 208 -2.699 -2 5 37 -38 $Cement #2 Front End Cap 18 208 -2.699 -2 31 36 -37 $Cement #2 Outer Shell 19 208 -2.699 -32 5 36 -37 $Cement #2 Inner Shell 20 228 -3.35 -31 32 36 -37 $Cement #2 21 457 -2.64 -2 5 38 -20 $Metamic #4 22 252 -11.35 -2 5 20 -21 $Lead #3 23 252 -11.35 -2 5 21 -22 $Lead #4

104

24 457 -2.64 -2 5 22 -23 $Metamic #5 25 208 -2.699 -2 5 23 -39 $Cement #3 Back End Cap 26 208 -2.699 -2 5 40 -41 $Cement #3 Front End Cap 27 208 -2.699 -2 31 39 -40 $Cement #3 Outer Shell 28 208 -2.699 -32 5 39 -40 $Cement #3 Inner Shell 29 228 -3.35 -31 32 39 -40 $Cement #3 30 457 -2.64 -2 5 41 -24 $Metamic #6 31 252 -11.35 -2 5 24 -25 $Lead #5 32 252 -11.35 -2 5 25 -26 $Lead #6 33 204 -0.001125 -2 5 26 -8 $Air Gap Inner 34 204 -0.001125 -32 7 -14 $1.5 in Hole, Rx to Bismuth 35 204 -0.001125 -5 6 -7 $1.25 in Hole, Al Rx End Cap 36 204 -0.001125 -5 14 -15 $1.25 in Hole, Metamic #1 37 204 -0.001125 -5 30 -9 $1.25 in Hole, Bismuth to Inner Collimator Front 38 208 -2.699 -1 5 6 -7 $Al Back End Cap, Inner 39 208 -2.699 -1 5 8 -9 $Al Front End Cap, Inner 40 208 -2.699 -1 2 7 -8 $Al Shell, Inner 41 457 -2.64 -4 5 11 -42 $Metamic #7 42 252 -11.35 -4 5 42 -43 $Lead #7 43 252 -11.35 -4 5 43 -44 $Lead #8 44 208 -2.699 -4 5 44 -49 $Cement #4 Back End Cap 45 208 -2.699 -4 5 50 -51 $Cement #4 Front End Cap 46 208 -2.699 -4 48 49 -50 $Cement #4 Outer Shell 47 208 -2.699 -32 5 49 -50 $Cement #4 Inner Shell 48 228 -3.35 -48 32 49 -50 $Cement #4 49 457 -2.64 -4 5 51 -45 $Metamic #8 50 252 -11.35 -4 5 45 -46 $Lead #9

105

51 252 -11.35 -4 5 46 -47 $Lead #10 52 204 -0.001125 -4 5 47 -12 $Air Gap Outer 53 204 -0.001125 -5 10 -13 $1.25 in Hole, Outer 54 208 -2.699 -3 5 10 -11 $Al Back End Cap, Outer 55 208 -2.699 -3 5 12 -13 $Al Front End Cap, Outer 56 208 -2.699 -3 4 11 -12 $Al Shell, Outer 57 204 -0.001125 -1 9 -10 $Air Gap, Shutter 58 228 -3.35 6 -10 53 -52 55 -54 1 $Biological Shielding #1 59 228 -3.35 10 -13 53 -52 55 -54 3 $Biological Shielding #2 60 204 -0.001125 -56 (-6 :13 :-53 :52 :-55 :54 ) $atmosphere 61 0 56 $void c SURFACE CARDS c Al Shell 1 cx 7.52 $Inner Al Pipe OD 2 cx 6.985 $Inner Al Pipe ID/Inner Aperture OD 3 cx 8.79 $Outer Al Pipe OD 4 cx 8.255 $Outer Al Pipe ID/Outer Aperture OD 5 cx 1.5875 $1.25 in Hole 6 px 0 $Al Cap, Rx end, Back 7 px 0.3175 $Al Cap, Rx end, Front/Graphite Back 8 px 103.8225 $Al Cap, Shutter side of Inner, Back 9 px 107.315 $Al Cap, Shutter side of Inner, Front 10 px 163.83 $Al Cap, Shutter side of Outer, Back 11 px 164.1475 $Al Cap, Shutter side of Outer, Front 12 px 200.9775 $Al Cap, BP end, Back 13 px 204.47 $Al Cap, BP end, Front

106 c Graphite Surfaces 14 px 13.0175 $Graphite Front c Metamic/Lead Surfaces Inner 15 px 13.6525 $Metamic #1 Front 16 px 45.085 $Metamic #2 Front 17 px 47.625 $Lead #1 Front 18 px 50.165 $Lead #2 Front 19 px 50.8 $Metamic #3 Front 20 px 71.755 $Metamic #4 Front 21 px 74.295 $Lead #3 Front 22 px 76.835 $Lead #4 Front 23 px 77.47 $Metamic #5 Front 24 px 98.425 $Metamic #6 Front 25 px 100.965 $Lead #5 Front 26 px 103.505 $Lead #6 Front c Bismuth Surfaces 27 cx 6.35 $Bismuth Shell ID/Bismuth OD 28 px 13.81125 $Bismuth Back Cap Inside 29 px 23.97125 $Bismuth Front Cap Inside 30 px 24.13 $Bismuth Front Cap End c Cement Surfaces, Inner 31 cx 6.6675 $Cement Outer Shell ID/Cement ID 32 cx 1.905 $Cement Inner Shell OD, Inner and Outer/Graphite Hole 33 px 24.4475 $#1 Cement Back End, Inside 34 px 44.1325 $#1 Cement Front End, Inside 35 px 44.45 $#1 Cement Front 36 px 51.1175 $#2 Cement Back End, Inside

107

37 px 70.8025 $#2 Cement Front End, Inside 38 px 71.12 $#2 Cement Front 39 px 77.7875 $#3 Cement Back End, Inside 40 px 97.4725 $#3 Cement Front End, Inside 41 px 97.79 $#3 Cement Front c Metamic/Lead Surfaces Outer 42 px 164.7825 $Metamic #7 Front 43 px 167.3225 $Lead #7 Front 44 px 169.8625 $Lead #8 Front 45 px 195.8975 $Metamic #8 Front 46 px 198.4375 $Lead #9 Front 47 px 200.66 $Lead #10 Front c Cement Surfaces, Outer 48 cx 7.9375 $Cement Outer Shell ID/Cement ID 49 px 170.18 $#4 Cement Back End, Inside 50 px 194.945 $#4 Cement Front End, Inside 51 px 195.2625 $#4 Cement Front c Concrete Surfaces 52 py 91.44 53 py -91.44 54 pz 91.44 55 pz -91.44 c Atmosphere 56 so 294 mode n m204 7014.70c -0.755636 $air (US S. Atm at sea level)

108

8016.70c -0.231475 18036.70c -3.9e-005 18038.70c -8e-006 18040.70c -0.012842 m236 6000.70c -1 $graphite m310 83209.70c -1 $Bismuth, m252 82206.70c -0.242902 $Lead 82207.70c -0.223827 82208.70c -0.53327 m456 1001.70c -0.143716 $Polyethylene, 6000.70c -0.856284 m457 5010.70c 4 $Metamic 6000.70c 1 13027.70c 1 m228 1001.70c 0.11 $Barytes Concrete 8016.70c 0.6 20000.62c 0.04 16032.70c 0.1 19000.62c 0.04 56138.70c 0.11 m208 13027.70c -1 $Aluminum imp:n 1 60r $ 1, 59

109

Appendix D: Additional weight and cost estimates

Table D-1: Beam Port 1, Inner Collimator estimated weight

Beam Port 1, Inner Collimator Al Tube Borated Graphite Bismuth Lead Metamic® Assembly Cement Quantity 1 1 1 3 6 4 Individual Mass 16.71 6.65 28.98 11.72 9.1 0.53 (lbs) Total Part Mass 16.71 6.65 28.98 35.17 54.58 2.12 (lbs) Collimator Total Total 127.48 144.19 (lbs) (lbs)

Table D-2: Beam Port 2, Inner Collimator estimated weight

Beam Port 2, Inner Collimator

Al Tube Borated Graphite Sapphire Bismuth Lead Metamic® Assembly Cement Quantity 1 1 1 1 3 4 4 Individual 17.62 4.75 11.24 28.98 11.72 9.1 0.53 Mass (lbs) Total Part Mass 17.62 4.75 11.24 28.98 35.17 36.39 2.12 (lbs) Collimator Total 118.63 136.25 Total (lbs) (lbs)

110

Table D-3: Outer Collimator estimated weight

Outer Collimator Al Tube Assembly Borated Cement Lead Metamic®

Quantity 1 1 4 2 Individual Mass (lbs) 8.71 20.48 12.94 0.75 Total Part Mass (lbs) 8.71 20.48 51.74 1.91 Collimator Total (lbs) 73.71 Total (lbs) 82.43

Table D-4: Rail/Lift system estimated cost excluding fabricated steel pieces: 1=PBC Linear, 2=McMaster-Carr, 3=Dayton/Joyce

Price Total Price Part Quantity ($) ($) 1 1.5" Shaft Rail Assembly, 36" long 2 397.36 794.72 1 1.5" Double Roller Pillow Block Bearing 4 391.97 1567.88 1 1" Single Flange Mount 4 85.54 342.16 1 1" Single Pillow Block Ball Bearing 4 64.12 256.48 1 1" Shaft, 24" Long 4 28.49 113.96

2 5172T350, Hi Load Metal Gear, 36 tooth, 3/4" Shaft 2 60.73 121.46

2 5172T360, Hi Load Metal Gear, 48 tooth, 7/8" Shaft 2 81.31 162.62

2 5174T23 Hi Load Metal Gear Rack, 1" wide x 1" high x 6' long 2 92.28 184.56

2 1497K620 3/4" Dia. Keyed Shaft, 36" long 1 42.88 42.88 2 1497K820 7/8" Dia. Keyed Shaft, 36" long 1 53.64 53.64 6494K130 Sealed Steel Bearing w/ Nickel Plated Cast Iron Housing, 2 4 34.74 138.96 3/4" Shaft

2 7728T540 Sealed Steel Bearing with Cast Iron Housing 2 83.43 166.86

2 6280K961 19 tooth Sprocket for 50 Roller Chain, 7/8" Shaft 1 36.73 36.73

2 2741T101 38 tooth Sprocket for 50 Roller Chain, 3/4" Shaft 1 65.32 65.32

2 6261K175 50 Roller Chain 5 5.98 29.9 2 47065T103 Single Rail, Silver, 1-1/2" Solid, 2 ft. long 3 18.19 54.57 Continued

111

Table D-4 continued 2 47065T103 Single Rail, Silver, 1-1/2" Solid, 1 ft. long 1 10.3 10.3 2 47065T843 Rail Mounting Foot, 1-1/2" 6 14.6 87.6 2 47065T12 Rail Brace, 1-1/2" x 6" long 2 16 32 2 47065T271 Rail Corner, 1-1/2" 2 8.77 17.54 2 47065T679 Rail Corner Gusset, 1-1/2" 2 7.58 15.16 2 47065T7620 Rail Corner Extended Gusset, 1-1/2" 2 11.92 23.84 47065T234 Drop in Fastener with Stud for 1-1/2" rail, 5/16"-18 thread x 2 4 1.58 6.32 31/32" long 3 Lift System Quote 1 2932.4 2932.4 Total 7257.86

112