WO2019200386A1 - Neutron shielding and absorption materials - Google Patents

Neutron shielding and absorption materials Download PDF

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Publication number
WO2019200386A1
WO2019200386A1 PCT/US2019/027492 US2019027492W WO2019200386A1 WO 2019200386 A1 WO2019200386 A1 WO 2019200386A1 US 2019027492 W US2019027492 W US 2019027492W WO 2019200386 A1 WO2019200386 A1 WO 2019200386A1
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isotopes
neutron
gadolinium
cadmium
materials
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PCT/US2019/027492
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French (fr)
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Danielle C. CASTLEY
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Trustees Of Dartmouth College
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Priority to US62/657,264 priority
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Publication of WO2019200386A1 publication Critical patent/WO2019200386A1/en

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    • GPHYSICS
    • G01MEASURING; TESTING
    • G01TMEASUREMENT OF NUCLEAR OR X-RADIATION
    • G01T3/00Measuring neutron radiation

Abstract

In some embodiments, the present disclosure pertains to a radiation absorbing composition. In some embodiments, the composition includes a base material and one or more isotopes associated with the base material. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns. In some embodiments, the present disclosure pertains to methods of absorbing neutrons from an environment by associating the environment with the radiation absorbing compositions of the present disclosure such that the association results in the absorption of neutrons from the environment onto the composition. Additional embodiments of the present disclosure pertain to methods of preparing the radiation absorbing compositions.

Description

TITLE

NEUTRON SHIELDING AND ABSORPTION MATERIALS CROSS-REFERENCE TO RELATED APPLICATIONS

[0001] This application claims priority to U.S. Provisional Patent Application No. 62/657,264, filed on April 13, 2018. The entirety of the aforementioned application is incorporated herein by reference.

BACKGROUND [0002] Radiation absorbing compositions are used in aerospace, space, medical, defense and military, scientific, and nuclear applications. Within numerous industries (e.g., the nuclear industry), there is a growing need for low density (e.g., < 2 g/cm ) and/or high temperature (e.g., > 200 °C) neutron shielding materials to improve the safety and reduce costs for numerous applications, such as new reactor designs and nuclear fuel management. The most common neutron shielding materials are boron or lithium containing polyethylene, polyamide composites, or water. Neutron shielding materials rely on slowing neutrons to a thermal state to increase the probability of absorption by boron.

[0003] As such, a need exists for new radiation absorbing compositions with improved properties that can be used to absorb and/or shield neutron radiation, such as higher thermal conductivity, higher thermal stability, lower density, and better gamma shielding. The present disclosure addresses the aforementioned needs, for example, by overcoming limitations associated with thermal, mechanical, and fabrication/geometric properties.

SUMMARY

[0004] In some embodiments, the present disclosure pertains to a radiation absorbing composition. In some embodiments, the composition includes a base material and one or more isotopes associated with the base material. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns.

[0005] In some embodiments, the present disclosure pertains to a method of absorbing neutrons from an environment. In some embodiments, the method includes associating the environment with a composition to result in the absorption of the neutrons from the environment onto the composition. In some embodiments, the composition includes a base material and one or more isotopes associated with the base material. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns.

[0006] In some embodiments, the present disclosure pertains to a method of preparing a radiation absorbing composition. In some embodiments, the method includes mixing a precursor material with one or more isotopes and a curing agent to result in the formation of a base material from the precursor material. In some embodiments, the base material becomes associated with the one or more isotopes. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 bams.

DESCRIPTION OF THE DRAWINGS

[0007] FIGURE 1A depicts a radiation absorbing composition according to an embodiment of the present disclosure.

[0008] FIGURE IB illustrates a method of absorbing neutrons from an environment according to an embodiment of the present disclosure. [0009] FIGURE 1C illustrates a method of preparing a radiation absorbing composition according to an embodiment of the present disclosure.

[0010] FIGURE 2 illustrates MCNP® model of neutron shielding material. The shielding ability of different materials was compared for effectiveness for thermal and 2 MeV neutrons. The material was modeled as a sphere with the detector on the surface of the shielding material to ensure an accurate neutron count. [0011] FIGURE 3 illustrates the R3 beam port at the Rhode Island Nuclear Science Center. Samples were placed in the carriage and the carriage is slid in front of the fast neutron beamline.

[0012] FIGURE 4 illustrates a schematic of the neutron attenuation test set-up in the beam port enclosure (not to scale). The source is the reactor and the beam is 12’ from the sample. A Cd plate is used when measuring the fast neutron macroscopic cross section as it stops the epithermal and thermal neutrons before entering the sample. Paraffin is placed between the sample and the BF-3 detector to moderate the neutrons enough for detection by the BF-3 detector.

[0013] FIGURE 5A illustrates a photo of the lead pig used for secondary gamma testing. The lead pig is 2 cm in diameter.

[0014] FIGURE 5B illustrates a schematic of the cross-sectional view of the sample in the lead pig lined with the radiachromic film.

[0015] FIGURE 6 illustrates a comparison of the neutron dose (bar graph) and photon dose rates (line graph) on the outer surface of each material composition from the MCNP® simulation. [0016] FIGURE 7 illustrates a cross sectional view of the (FIG. 7A) Epoxy, neat and (FIG. 7B) epoxy with 10 wt% Gd203 and 2 wt% B4C

[0017] FIGURE 8 illustrates cross sections of the samples with 3 wt% of the additives (FIG. 8A) B4C, (FIG. 8B) Gd203, and (FIG. 8C) Sm203. At 3 wt% the additives were evenly distributed. [0018] FIGURE 9 illustrates Cross sections of the samples with 9 wt% of the additives (FIG.

9A) B4C, (FIG. 9B) Gd203, and (FIG. 9C) Sm203. The B4C in the 9 wt% sample was evenly distributed, uniformly darkened the epoxy, and the individual nanoparticles were not visible.

[0019] FIGURE 10 illustrates cross sections of the samples with 27 wt% neutron absorber (FIG. 10A) 27 wt% B4C, (FIG. 10B) 27 wt% Gd203, and (FIG. 10C) 27 wt% Sm203. The B4C was not homogeneously distributed at 27 wt% loading in the epoxy but the Gd203 and Sm203 were evenly distributed.

[0020] FIGURE 11A illustrates TGA Curve for all samples matched the degradation of the epoxy curve (shown) except the 27 wt% B4C sample due to the challenges in evenly distributing the nanoparticles.

[0021] FIGURE 11B illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% B4C in epoxy.

[0022] FIGURE 11C illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% Gd203 in epoxy.

[0023] FIGURE 11D illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% Sm203 in epoxy

[0024] FIGURE 12 illustrates relative neutron transmission vs. wt% neutron absorber for 2 MeV beamline. Samples were placed in front of the beamline and the number of neutrons with and without the samples was measured. The ratio Flo indicates the neutrons that were transmitted through the sample. [0025] FIGURE 13 illustrates relative neutron transmission of a 2 MeV beamline through 3” thick samples containing neutron absorbing additives and graphene platelets.

[0026] FIGURE 14 illustrates experimental photon absorbance to the radiachromic film from the varying material combinations.

DETAIUED DESCRIPTION [0027] It is to be understood that both the foregoing general description and the following detailed description are illustrative and explanatory, and are not restrictive of the subject matter, as claimed. In this application, the use of the singular includes the plural, the word“a” or“an” means“at least one”, and the use of“or” means“and/or”, unless specifically stated otherwise. Furthermore, the use of the term“including”, as well as other forms, such as“includes” and “included”, is not limiting. Also, terms such as“element” or“component” encompass both elements or components comprising one unit and elements or components that include more than one unit unless specifically stated otherwise.

[0028] The section headings used herein are for organizational purposes and are not to be construed as limiting the subject matter described. All documents, or portions of documents, cited in this application, including, but not limited to, patents, patent applications, articles, books, and treatises, are hereby expressly incorporated herein by reference in their entirety for any purpose. In the event that one or more of the incorporated literature and similar materials defines a term in a manner that contradicts the definition of that term in this application, this application controls. [0029] Neutron shielding and absorption materials prevent criticality in various nuclear applications and increase the safety of medical, aerospace, and space applications. For example, in nuclear applications, spent fuel assemblies are taken out from an atomic reactor, stored in water-cooled pools at the atomic power plant site for a preset time period to attenuate radiation dose and calorific power, and then transported to a storage facility (e.g., dry storage facility) or a processing facility (e.g., fuel reprocessing factory). A specially designed container, often referred to as a cask, is used to store and/or carry the spent nuclear fuel assembly.

[0030] There are generally various types of casks, such as, but not limited to, transfer casks, transport casks, storage casks, and dual-purpose storage and transport casks. Typically, transfer casks are designed to be lighter than storage casks because a transfer cask must be lifted, handled, and transported by, for example, a crane, or other machinery.

[0031] Current neutron shielding and absorption materials have limited thermal properties, are susceptible to degradation from secondary gamma radiation such as that caused when isotopes capture neutrons, and, generally, neutron absorbers have a relatively high density. Such properties pose concerns for long-term fuel management. [0032] For example, borated polyethylene, water, and concrete are the most commonly used neutron shielding materials in spent fuel storage applications. Concrete is multi-purpose in shielding neutron radiation, shielding gamma radiation, and providing structural support or protection from impact. Concrete is stable at high temperatures but has a relatively high density.

[0033] Conversely, water has a lower density but begins to boil at 100 °C, and therefore requires cumbersome mechanical cooling systems. The neutron attenuation abilities of both water and concrete have been leveraged with neutron absorbing additives such as boric acid, boron carbide, and ferro-boron. However, the densities and/or max operating temperatures of such materials are still limiting.

[0034] Accordingly, a need exists for more effective radiation absorption compositions and methods. Various embodiments of the present disclosure address the aforementioned needs. [0035] In some embodiments, the present disclosure pertains to radiation absorbing compositions that include a base material and one or more isotopes. In some embodiments illustrated in FIG. 1A, the radiation absorbing compositions of the present disclosure have a conformation 10, which includes a base material 12 and one or more isotopes 14 and 16 associated with the base material 12. In some embodiments, the one or more isotopes 14 and 16 have an individual or combined thermal neutron cross section of more than 50 barns. In some embodiments, the one or more isotopes 14 and 16 can be two or more different isotopes of a same element. In some embodiments, the one or more isotopes 14 and 16 can be two or more different isotopes of different elements. In some embodiments, the one or more isotopes 14 and 16 are part of one or more neutron absorbing compounds. In some embodiments, the base material 12 can additionally include additives (not shown).

[0036] Additional embodiments of the present disclosure pertain to methods of absorbing neutrons from an environment. In some embodiments illustrated in FIG. IB, the methods of the present disclosure include associating the environment with a radiation absorbing composition of the present disclosure (step 20) to result in the absorption of neutrons from the environment onto the composition (step 22). In some embodiments, the radiation absorbing composition can include a base material and one or more isotopes associated with the base material. In some embodiments, the one or more isotopes can be two or more different isotopes of a same element.

In some embodiments, the one or more isotopes can be two or more different isotopes of different elements. In some embodiments, the one or more isotopes are part of one or more neutron absorbing compounds.

[0037] Further embodiments of the present disclosure pertain to methods of preparing the radiation absorbing compositions of the present disclosure. In some embodiments illustrated in FIG. 1C, the method involves mixing a precursor material with one or more isotopes of the present disclosure (step 30) and a curing agent (step 32) to result in the formation of a base material from the precursor material (step 34) and the formation of the radiation absorbing composition (step 36), where the base material is associated with one or more isotopes. In some embodiments, the one or more isotopes can be two or more different isotopes of a same element. In some embodiments, the one or more isotopes can be two or more different isotopes of different elements. In some embodiments, the one or more isotopes are part of one or more neutron absorbing compounds.

[0038] In some embodiments, the methods of the present disclosure can also include a step of mixing additives (step 38) before mixing the curing agent (step 32). In some embodiments, the methods of the present disclosure can also include a step of mixing additives (step 40) after mixing the curing agent.

[0039] As set forth in more detail herein, the methods and compositions of the present disclosure can have numerous embodiments. For instance, the radiation absorbing compositions of the present disclosure can include various base materials, isotopes, and additives. Furthermore, the one or more isotopes can be derived from one or more elements. In some embodiments, the one or more isotopes can be part of one or more neutron absorbing compounds.

[0040] Furthermore, various methods may be utilized to absorb neutrons from an environment by using the radiation absorbing compositions of the present disclosure. Various methods may also be utilized to prepare the radiation absorbing compositions of the present disclosure. [0041] Radiation Absorbing Compositions [0042] As set forth in more detail herein, the radiation absorbing compositions of the present disclosure can include various base materials, one or more isotopes, and additives. In some embodiments, the one or more isotopes are derived from one or more elements. In some embodiments, the one or more isotopes can be part of one or more neutron absorbing compounds. In some embodiments, the one or more isotopes can include two or more different isotopes of a same element. In some embodiments, the one or more isotopes can include two or more different isotopes of different elements.

[0043] Moreover, the radiation absorbing compositions of the present disclosure may be associated with an environment in various manners to absorb neutrons. In addition, the radiation absorbing compositions may have various advantageous properties.

[0044] Base Material

[0045] The radiation absorbing compositions of the present disclosure can include various types of base materials. For instance, in some embodiments, the base material can be a polymer matrix, an epoxy, ceramics, composites, fiber-enforced composites, fiber-enforced polymers, plastics, cements, graphites, rubber, perlite, colemanite, paraffin, phenolic resin, wood, concrete, aluminum, steel, or combinations thereof.

[0046] In some embodiments, the base material can be a polymer matrix. In some embodiments, the polymer matrix can be silicone, room temperature vulcanizing (RTV) silicone, polydimethylsiloxane (PDMS), thermoset polymers, or combinations thereof. [0047] In some embodiments, the base material can be an epoxy. In some embodiments, the epoxy can be a silicone rubber, a bisphenol A epoxy, a novolak epoxy, an alicyclic glycidyl ether-type epoxy, a glycidyl ester-type epoxy, a glycidyl amine-type epoxy, a biphenol-type epoxy, or combinations thereof.

[0048] In particular embodiments, the base material can be a base thermoset polymer, such as, for example, Dow Chemical XIAMETER® RTV-3010 or Dow Chemical XIAMETER® RTV- 3120. In some embodiments, the base materials of the present disclosure exclude cement. The use of additional base materials in the radiation absorbing compositions of the present disclosure can also be envisioned.

[0049] Isotopes and Neutron Absorbing Compounds

[0050] The radiation absorbing compositions of the present disclosure can include various types of one or more isotopes associated with the base material. For instance, in some embodiments, the one or more isotopes comprise two or more different isotopes of a same element, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns. In some embodiments, the one or more isotopes comprise two or more different isotopes of different elements, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns. In some embodiments, the one or more isotopes can be derived from one or more elements, such as, but not limited to, a metal, a metalloid, a transition metal, a post transition metal, a lanthanide, an oxide, and a ceramic.

[0051] In some embodiments, the one or more isotopes are part of one or more neutron absorbing compounds. For instance, in some embodiments, the one or more neutron absorbing compounds can be boron-containing materials, gadolinium-containing materials, samarium- containing materials, cadmium-containing materials, molybdenum-containing materials, hafnium-containing materials, titanium-containing materials, dysprosium-containing materials, iron-containing materials, lithium-containing materials, ytterbium-containing materials, zinc- containing materials, or combinations thereof. In some embodiments, the one or more neutron absorbing compounds can include a plurality of different neutron absorbing compounds.

[0052] In some embodiments, the one or more neutron absorbing compounds can include at least one of boron carbide, boric acid, boron nitride, boron oxide, gadolinium oxide, gadolinium acetate, samarium oxide, cadmium oxide, molybdenum boride, hafnium diboride, titanium diboride, dysprosium titanate, lithium, gadolinium titanate, iron oxide, or combinations thereof. [0053] In some embodiments, the one or more neutron absorbing compounds can include a gadolinium-containing material and a boron containing material. In some embodiments, the one or more neutron absorbing compounds can be gadolinium oxide and boron carbide. In some embodiments, the one or more neutron absorbing compounds are in the form of nanoparticles.

[0054] In some embodiments, the one or more neutron absorbing compounds constitute from about 0.1 wt% to about 60 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds constitute from about 1 wt% to about 25 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds constitute from about 5 wt% to about 20 wt% by combined weight of the base material and one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure.

[0055] In some embodiments, the one or more neutron absorbing compounds can include boron carbide (B4C) in an amount of about 3 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include B4C in an amount of about 9 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include B4C in an amount of about 27 wt% by combined weight of the base material and one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure.

[0056] In some embodiments, the one or more neutron absorbing compounds can include gadolinium oxide (Gd203) in an amount of about 3 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include Gd203 in an amount of about 9 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include Gd203 in an amount of about 27 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure.

[0057] In some embodiments, the one or more neutron absorbing compounds can include samarium oxide (Sm203) in an amount of about 3 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include Sm203 in an amount of about 9 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure. In some embodiments, the one or more neutron absorbing compounds can include Sm203 in an amount of about 27 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions of the present disclosure.

[0058] In some embodiments, the one or more neutron absorbing compounds can be a plurality of neutron absorbing compounds that include at least two different neutron absorbing compounds. In some embodiments, the two different neutron absorbing compounds can include a gadolinium-containing material and a boron-containing material. In some embodiments, the two different neutron absorbing compounds can be Gd203 and B4C. In some embodiments, the two different neutron absorbing compounds can be Gd203 in an amount of about 10 wt% and B4C in an amount of about 2 wt% by combined weight of the base materials and the two different neutron absorbing compounds in the radiation absorbing compositions of the present disclosure.

[0059] In some embodiments, the two different neutron absorbing compounds can include a samarium-containing material and a boron containing material. In some embodiments, the two different neutron absorbing compounds can be Sm203 and B4C in the base material. In some embodiments, the base material can include more than two neutron absorbing compounds, for example, B4C, Gd203, and Sm203.

[0060] The neutron absorbing compounds of the present disclosure can be in various forms. For instance, in some embodiments, the neutron absorbing compounds are in the form of nanoparticles. In some embodiments, the nanoparticles have sizes that range from about 1 nm to about 500 nm in diameter. In some embodiments, the nanoparticles have sizes that range from about 1 nm to about 250 nm in diameter. In some embodiments, the nanoparticles have sizes that range from about 1 nm to about 100 nm in diameter. In some embodiments, the nanoparticles have sizes that are less than about 100 nm in diameter.

[0061] Neutron absorbing compounds may be associated with base materials in various manners. For instance, in some embodiments, the neutron absorbing compounds are uniformly dispersed throughout the base material. In some embodiments, the neutron absorbing compounds are intertwined with the base material. In some embodiments, the neutron absorbing compounds are positioned within the surface of a base material. In some embodiments, the neutron absorbing compounds are positioned within internal cavities of a base material. In some embodiments, the neutron absorbing compounds are evenly distributed throughout the base material. In some embodiments, the neutron absorbing compounds can be intertwined throughout a compound, element, composition, material, or combinations thereof. [0062] In some embodiments, the neutron absorbing compound can be a layered compound. In some embodiments, the neutron absorbing compound can be layered within compounds, elements, compositions, materials, or combinations thereof. In some embodiments, the neutron absorbing compound can be threaded through compounds, elements, compositions, materials, or combinations thereof. [0063] Isotope Elements

[0064] The one or more isotopes of the radiation absorbing compositions of the present disclosure can each include various types of isotopes. In some embodiments, the isotopes can be derived from one or more elements. In some embodiments, the isotopes can be different isotopes of the same element. In some embodiments, the isotopes can be different isotopes of different elements. In some embodiments, the isotopes can be one or more elements.

[0065] In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 bams. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 100 bams. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 150 bams. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 200 bams. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 250 bams. In some embodiments, the one or more isotopes have an individual or combined thermal neutron cross section of more than 300 barns.

[0066] The one or more isotopes of the present disclosure can be derived from various elements. For instance, in some embodiments, the element can be a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, or a ceramic. In some embodiments, the element can be boron, gadolinium, cadmium, samarium, lithium, hafnium, cobalt, titanium, dysprosium, erbium, europium, molybdenum, ytterbium, zinc, or iron.

[0067] The elements can have various isotopes. For instance, in some embodiments, the one or more isotopes can be natural or enriched isotopes. In some embodiments, the one or more isotopes can include, without limitation, lithium-6 (6Li), lithium-7 (7Li), boron- 10 (10B), boron- 11 (nB), gadolinium- 157 (157Gd), gadolinium- 152 (152Gd), gadolinium- 154 (154Gd), gadolinium- 155 (155Gd), gadolinium- 156 (156Gd), gadolinium- 158 (158Gd), gadolinium- 160 (160Gd), gadolinium- 185 (185Gd), samarium-l5l (151Sm), samarium-l49 (149Sm), samarium-l44 (144Sm), samarium-l50 (150Sm), samarium-l52 (152Sm), samarium-l54 (154Sm), cadmium-H3 (113Cd), cadmium-H6 (116Cd), cadmium-l06 (106Cd), cadmium-l08 (108Cd), cadmium-H4 (114Cd), cadmium-l lO (110Cd), cadmium-l l l (mCd), cadmium-H2 (112Cd), cadmium-l09 (109Cd), cadmium-H5 (115Cd), and cadmium-H7 (117Cd).

[0068] In some embodiments, the one or more isotopes mitigate negative effects of neutron absorption. In some embodiments, the negative effects can be secondary gamma production by the radiation absorbing composition after neutron absorption, alpha production by the radiation absorbing composition after neutron absorption, secondary x-ray production by the radiation absorbing composition after neutron absorption, and combinations thereof. In some embodiments, the one or more isotopes are uniformly dispersed throughout a base material. [0069] Additives

[0070] In some embodiments, the radiation absorbing compositions of the present disclosure can include various types of additives. For instance, in some embodiments, the additives can include, without limitation, fullerenes, copper nanomaterials, silver nanomaterials, aluminum nanomaterials, metal hydride, hydrogen-absorbing alloys, carbon allotropes, silicon carbides, conductive metals, iron, silicon, carbon, and oxygen, or combinations thereof.

[0071] In some embodiments, the radiation absorbing compositions can include a carbon allotrope. In some embodiments, the carbon allotrope can be graphene, graphene platelets, carbon nanotubes, carbon nanofibers, or combinations thereof. In some embodiments, the additives can include, without limitation, iron, silicon, carbon, and oxygen. In some embodiments, the additives can be present in the base material, associated with the one or more isotopes, present in the neutron absorbing composition, or combinations thereof.

[0072] In some embodiments, the additives are in the form of nanoparticles. In some embodiments, the additives are uniformly dispersed throughout the base material. In some embodiments, the additives are uniformly dispersed throughout the neutron absorbing compositions. In some embodiments, the additives are uniformly dispersed throughout the base material and the neutron absorbing compositions. In some embodiments, the additives are applied to an outer surface of the radiation absorbing compositions, for example, by chemical vapor deposition or coating. [0073] The additives of the present disclosure can have various effects on the compositions of the present disclosure. For instance, in some embodiments, the additives facilitate the attenuation of gamma radiation from the radiation absorbing compositions of the present disclosure. In some embodiments, the additives protect the radiation absorbing compositions from gamma radiation. In some embodiments, the additives mitigate secondary gamma radiation resulting from neutron absorption by the radiation absorbing compositions. In some embodiments, the additives provide for a high coefficient of thermal conductivity. In some embodiments, the additives can be utilized to improve electromagnetic frequency (EMF) shielding. [0074] Characteristics and Properties

[0075] The radiation absorbing compositions of the present disclosure can have various advantageous characteristics and properties. For instance, in some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at higher temperatures that are not significantly different from their neutron absorbing capabilities at lower temperatures. For instance, in some embodiments, the neutron absorbing capabilities of the compositions of the present disclosure at higher temperatures are not less than 90% of their neutron absorbing capabilities at lower temperatures. In some embodiments, the neutron absorbing capabilities of the compositions of the present disclosure at higher temperatures are not less than 95% of their neutron absorbing capabilities at lower temperatures.

[0076] The radiation absorbing compositions of the present disclosure can demonstrate the aforementioned neutron absorbing capabilities at various temperature ranges. For instance, in some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures at or above 180 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at or above 200 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at or above 250 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures at or above 300 °C.

[0077] In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures from about 180 °C to about 200 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures from about 200 °C to about 250 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures from about 225 °C to about 275 °C. In some embodiments, the radiation absorbing composition have neutron absorbing capabilities at temperatures from about 250 °C to about 300 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures from about 275 °C to about 325 °C. In some embodiments, the radiation absorbing compositions have neutron absorbing capabilities at temperatures from about 300 °C to about 350 °C. [0078] The radiation absorbing compositions of the present disclosure can also have various densities. For instance, in some embodiments, the radiation absorbing compositions have a density at or below 8 g/cm . In some embodiments, the radiation absorbing compositions have a y

density at or below 7.5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 7 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 6.5 g/cm3. In some embodiments, the radiation absorbing compositions have a density at or below 6 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 5.5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 4.5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 4 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 3.5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 3 g/cm3. In some embodiments, the radiation absorbing compositions have a density at or below 2.5 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 2 g/cm . In some embodiments, the radiation absorbing compositions have a density at or below 1.5 g/cm .

[0079] In some embodiments, the radiation absorbing compositions have a density from about 1 g/cm 3 to about 1.2 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 1.2 g/cm 3 to about 1.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 1.4 g/cm 3 to about 1.6 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 1.6 g/cm 3 to about 1.8 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 1.8 g/cm to about 2 g/cm . In some embodiments, the radiation absorbing compositions have a density from about 2 g/cm 3 to about 2.2 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 2.2 g/cm 3 to about 2.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 2.4 g/cm 3 to about 3.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 3.4 g/cm to about

4.4 g/cm . In some embodiments, the radiation absorbing compositions have a density from about 4.4 g/cm 3 to about 5.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 5.4 g/cm 3 to about 6.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 6.4 g/cm 3 to about 7.4 g/cm 3. In some embodiments, the radiation absorbing compositions have a density from about 7.4 g/cm to about 8 g/cm3. [0080] In some embodiments, the radiation absorbing compositions of the present disclosure have low hydrogen contents. For instance, in some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 15% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 10% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 5% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 4.5% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 4% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 3.5% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the neutron absorbing compositions are the one or more isotopes. [0081] In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 3.5% to about 4% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content of less than about 4% to about 4.5% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the radiation absorbing compositions have a hydrogen content less than about 4.5% to about 5% by combined weight of the base material and the neutron absorbing compositions. In some embodiments, the neutron absorbing compositions are the one or more isotopes. [0082] The radiation absorbing compositions of the present disclosure can have various neutron cross sections. For instance, in some embodiments, the radiation absorbing compositions have an effective neutron cross section of from about 0.019 cm 1 to about 0.31 cm 1. In some embodiments, the radiation absorbing compositions have an effective neutron cross section of from about 0.019 cm 1 to about 0.03 cm 1. In some embodiments, the radiation absorbing compositions have an effective neutron cross section of from about 0.02 cm 1 to about 0.03 cm 1. In particular embodiments, the effective neutron cross section can be 0.021 cm 1 for radiation absorbing compositions containing 9 wt% B4C, 0.031 cm 1 for radiation absorbing compositions containing 27 wt% Gd203, and 0.030 cm 1 for radiation absorbing compositions containing 3% Sm203. In a particular embodiment, the effective neutron cross section can be 0.026 cm 1 for radiation absorbing compositions containing 10 wt% Gd203 and 2 wt% B4C.

[0083] The radiation absorbing compositions of the present disclosure can also have various neutron resistance values. For instance, in some embodiments, the radiation absorbing compositions have a neutron resistance greater than about 1.0x10 n/cm . In some embodiments, the radiation absorbing compositions have a neutron resistance greater than about 1.0x10 n/cm . In some embodiments, the radiation absorbing compositions have a neutron resistance greater than about l.OxlO16 n/cm2. In some embodiments, the radiation absorbing compositions have a neutron resistance greater than about 1.0x10 n/cm . In some embodiments, the radiation absorbing compositions have a neutron resistance greater than about 1.0x10 n/cm . In some embodiments, the radiation absorbing compositions have a neutron resistance greater than about 1.0x10 19 n/cm 2. In some embodiments, the radiation absorbing compositions have a neutron resistance greater than 1.0x1020 n/cm 2.

[0084] The radiation absorbing compositions of the present disclosure can also have various thermal conductivities. For instance, in some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.1 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.2 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.3 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.4 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.5 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.6 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.7 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.8 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 0.9 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 1.0 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 1.2 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 1.4 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 1.6 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 1.8 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 2.0 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 2.2 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 2.4 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 2.6 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 2.8 W/mK. In some embodiments, the radiation absorbing compositions have a thermal conductivity above about 3 W/mK. [0085] In some embodiments, the radiation absorbing compositions of the present disclosure mitigate secondary gamma radiation that is emitted from the compositions after neutron absorption. For instance, in some embodiments, the radiation absorbing compositions mitigate secondary gamma radiation emitted from the compositions after neutron absorption. In some embodiments, the radiation absorbing compositions of the present disclosure mitigate negative effects of neutron absorption, such as, but not limited to, secondary gamma production by the radiation absorbing composition after neutron absorption, alpha production by the radiation absorbing composition after neutron absorption, secondary x-ray production by the radiation absorbing composition after neutron absorption, and combinations thereof.

[0086] Absorbing Neutrons from an Environment [0087] Additional embodiments of the present disclosure pertain to methods of absorbing neutrons from an environment. Such methods generally involve associating the environment with the compositions of the present disclosure to result in the absorption of the neutrons from the environment onto the composition. As set forth in more detail herein, the radiation absorbing compositions of the present disclosure may be utilized to absorb neutrons from various environments in various manners.

[0088] Association of an Environment with Neutron Absorbing Compositions

[0089] The association of an environment with radiation absorbing compositions of the present disclosure can be accomplished in various manners. For instance, in some embodiments, the association can occur when the radiation absorbing compositions are in proximity to a radiation- producing facility, such as, but not limited to, an atmosphere of a nuclear power plant. In some embodiments, the association can occur when the radiation absorbing compositions are used to create or cover various containers for use in high radiation applications. For instance, in some embodiments, the containers are casks, such as, but not limited to, casks for storing or transporting radioactive material. In some embodiments, the containers may be nuclear fuel transportation casks. In some embodiments, the containers may be storage casks. In some embodiments, the containers may be dual-purpose storage and transfer casks.

[0090] In some embodiments, the casks can include an inner shell and an outer shell. In some embodiments, the space between the inner shell and the outer shell can include the radiation absorbing compositions of the present disclosure. In some embodiments, the inner shell, the outer shell, or both the inner and outer shell include the radiation absorbing compositions.

[0091] In some embodiments, the present disclosure provides equipment or an equipment housing including the radiation absorbing compositions disclosed herein such that the equipment or equipment housing come in contact with neutron radiation. In some embodiments, the equipment or equipment housing is applicable to the space or aerospace industry. For example, in some embodiments, the equipment may be a space satellite, a space system, or combinations thereof. In some embodiments, the equipment or equipment housing is for the medical industry and can include, but is not limited to, shielding for radioactive testing. [0092] In some embodiments, environments can become associated with the radiation absorbing compositions of the present disclosure by being in general proximity to a radiation source. In some embodiments, environments can become associated with the radiation absorbing compositions of the present disclosure by bombardment of a radiation beam. In some embodiments, environments can become associated with the radiation absorbing compositions of the present disclosure by being used for walls, linings, liners, insulators, shielding devices, or combinations thereof, which are in contact with, or surrounded by, radiation. In some embodiments, the radiation absorbing compositions of the present disclosure become associated with the environment by flowing the environment through the composition. In some embodiments, the environment is an atmosphere of a nuclear power plant. In some embodiments, the environment is due to activities cause by aerospace, space, medical, defense and military, scientific, and nuclear applications.

[0093] Absorbing Neutrons from the Environment

[0094] The absorption of neutrons from an environment by the radiation absorbing compositions of the present disclosure can be accomplished in various manners. For instance, in some embodiments, associating the environment with the radiation absorbing compositions of the present disclosure can result in the absorption of neutrons from the environment onto the composition.

[0095] In some embodiments, neutrons are absorbed from the environment onto the composition by being absorbed onto a surface of the composition. In some embodiments, neutrons are absorbed from the environment onto the composition by being absorbed into the composition.

[0096] In some embodiments, absorption of neutrons from the environment is based, at least in part, on a relationship between neutron absorber particle size of the neutron absorbing material and neutron attenuation at a same weight percentage of the neutron absorbing material. In some embodiments, attenuation improves as neutron absorbing particle size of the neutron absorbing material decreases. [0097] In some embodiments, as neutron absorbing material weight percent increases, neutron transmission decreases. In some embodiments where the neutron absorbing compositions include submicron nanoparticles, the neutron transmission for 2% neutron absorbing material loading is even less than neutron transmission for one-micron neutron absorbing compositions particles at 5% neutron absorbing material loading. As such, in some embodiments, submicron neutron absorbing material nanoparticles can be used for neutron shielding and absorption. In some embodiments, neutron absorption is optimized with evenly distributed particles of the neutron absorbing compositions.

[0098] In some embodiments, after a neutron has been absorbed in the radiation absorbing composition, the method can be repeated and/or continued. In some embodiments, the method is repeated and/or continued until there are no more neutrons left in the environment. In some embodiments, the method is repeated and/or continued until the radiation absorbing composition is fully utilized. In some embodiments, the method is repeated and/or continued until the radiation absorbing composition is replaced with a new radiation absorbing composition. In some embodiments, a new radiation absorbing composition is used when an existing radiation absorbing composition needs to be replaced, for example, during routine maintenance at a nuclear waste storage facility.

[0099] Methods of Preparing Radiation Absorbing Compositions

[00100] Additional embodiments of the present disclosure pertain to methods of preparing the radiation absorbing compositions of the present disclosure. Such methods generally include mixing a precursor material with one or more isotopes and a curing agent to result in the formation of a base material from the precursor material that is associated with the one or more isotopes. In some embodiments, the methods of preparing the radiation absorbing compositions of the present disclosure can include mixing the precursor material, the one or more isotopes, and the curing agent with an additive.

[00101] In some embodiments, the mixing of the precursor materials, one or more isotopes, curing agents, and additives can occur simultaneously or in various orders. For instance, in some embodiments illustrated in FIG. 1C, a precursor material may be mixed with one or more isotopes (step 30) before a curing agent is added (step 32). In some embodiments, an additive may also be added before the curing agent is added (step 38). In some embodiments, an additive may be added after the curing agent is added (step 40).

[00102] As set forth in more detail herein, the methods of the present disclosure can utilize various precursor materials, one or more isotopes, curing agents, and additives. Moreover, the aforementioned components may be mixed in various manners to form the compositions of the present disclosure.

[00103] Precursor Materials

[00104] The radiation absorbing compositions of the present disclosure can be prepared using various types of precursor materials. For instance, in some embodiments, the precursor materials are materials that form the base materials of the present disclosure after curing. In some embodiments, the precursor materials can be a silicone rubber, a bisphenol A epoxy, a novolak epoxy, an alicyclic glycidyl ether-type epoxy, a glycidyl ester-type epoxy, a glycidyl amine-type epoxy, a biphenol-type epoxy, or combinations thereof. In some embodiments, the precursor materials can be applied onto a surface or mold using three-dimensional (3D) printing. In some embodiments, the radiation absorbing compositions of the present disclosure can be formed by 3D printing using the precursor materials described herein.

[00105] Curing Agents

[00106] The radiation absorbing compositions of the present disclosure can be prepared using various types of curing agents. For instance, in some embodiments, the curing agents can include, without limitation, an amine-type hardener, an acid anhydrate-type hardener, an imidazole-type hardening promoter, water, or combinations thereof. In some embodiments, the curing agents can be an amine-type hardener such as, but not limited to, an aromatic amine, an alicyclic amine, a polyamide amine, or combinations thereof. In some embodiments, the curing agents can be applied onto a surface or mold using 3D printing. In some embodiments, the radiation absorbing compositions of the present disclosure can be formed by 3D printing using the curing agents described herein. [00107] Isotopes and Neutron Absorbing Compounds

[00108] The radiation absorbing compositions of the present disclosure can be prepared using various types of one or more isotopes. For instance, in some embodiments, the one or more isotopes include two or more different isotopes of a same element, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns. In some embodiments, the one or more isotopes include two or more different isotopes of different elements, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns. In some embodiments, the one or more isotopes are derived from one or more elements, such as, but not limited to, a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, or a ceramic. In some embodiments, the one or more isotopes can be applied onto a surface or mold using 3D printing. In some embodiments, the radiation absorbing compositions of the present disclosure can be formed by 3D printing using the one or more isotopes described herein.

[00109] In some embodiments, the one or more isotopes are part of one or more neutron absorbing compounds. In some embodiments, the one or more neutron absorbing compounds discussed above can be utilized, and can include, without limitation, boron-containing materials, gadolinium-containing materials, samarium-containing materials, cadmium-containing materials, molybdenum-containing materials, hafnium-containing materials, titanium-containing materials, dysprosium-containing materials, iron-containing materials, lithium-containing materials, ytterbium-containing materials, zinc-containing materials, or combinations thereof.

[00110] In some embodiments, the radiation absorbing compositions can include a plurality of neutron absorbing compositions. In some embodiments, the neutron absorbing compositions can include boron carbide, boric acid, boron nitride, boron oxide, gadolinium oxide, gadolinium acetate, samarium oxide, cadmium oxide, molybdenum boride, hafnium diboride, titanium diboride, dysprosium titanate, lithium, gadolinium titanate, iron oxide, or combinations thereof.

[00111] In some embodiments, the one or more neutron absorbing compounds can include a plurality of different neutron absorbing compounds. In some embodiments, the neutron absorbing compounds can be prepared using at least two different neutron absorbing compounds. In some embodiments, the one or more neutron absorbing compounds can include a gadolinium- containing material and a boron containing material. In some embodiments, the one or more neutron absorbing compounds can be gadolinium oxide and boron carbide. In some embodiments, the one or more neutron absorbing compounds are in the form of nanoparticles. In some embodiments, the one or more neutron absorbing compounds constitute from about 0.1 wt% to about 60 wt% by combined weight of the base material and the one or more neutron absorbing compounds in the radiation absorbing compositions. In some embodiments, the neutron absorbing compounds can be applied onto a surface or mold using 3D printing. In some embodiments, the radiation absorbing compositions of the present disclosure can be formed by 3D printing using the neutron absorbing compounds described herein.

[00112] In some embodiments described above, each of the two different neutron absorbing compounds can include different isotopes of an element. For example, in some embodiments, the element can be a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, and a ceramic. In some embodiments, the element can be boron, gadolinium, cadmium, samarium, lithium, hafnium, cobalt, titanium, dysprosium, erbium, europium, molybdenum, ytterbium, zinc, or iron.

[00113] In some embodiments, the elements can have various isotopes. For instance, in some embodiments, the isotopes can be natural or enriched isotopes, such as, but not limited to, lithium-6 (6Li), lithium-7 (7Li), boron-lO (10B), boron-l l (nB), gadolinium- 157 (157Gd), gadolinium- 152 (152Gd), gadolinium- 154 (154Gd), gadolinium- 155 (155Gd), gadolinium- 156

(156Gd), gadolinium- 158 (158Gd), gadolinium- 160 (160Gd), gadolinium- 185 (185Gd), samarium-

151 (151Sm), samarium-l49 (149Sm), samarium-l44 (144Sm), samarium-l50 (150Sm), samarium-

152 (152Sm), samarium-l54 (154Sm), cadmium-H3 (113Cd), cadmium-H6 (116Cd), cadmium-l06 (106Cd), cadmium-l08 (108Cd), cadmium-H4 (114Cd), cadmium-l lO (110Cd), cadmium-l l l (mCd), cadmium-H2 (112Cd), cadmium-l09 (109Cd), cadmium-H5 (115Cd), and cadmium-H7 ( Cd). In some embodiments, the elements can provide the one or more isotopes. In some embodiments, the one or more isotopes are uniformly dispersed throughout the base material.

[00114] Additives [00115] The radiation absorbing compositions of the present disclosure can be prepared using various types of additives. For instance, in some embodiments, the additives discussed above can be utilized, and can include, without limitation, fullerenes, copper nanomaterials, silver nanomaterials, aluminum nanomaterials, metal hydride, hydrogen-absorbing alloys, carbon allotropes, silicon carbides, conductive metals, iron, silicon, carbon, , oxygen, or combinations thereof. In some embodiments, the additives can include a carbon allotrope. In some embodiments, the carbon allotrope can be graphene, graphene platelets, carbon nanotubes, carbon nanofibers, or combinations thereof. In some embodiments, the additives can include, without limitation, iron, silicon, carbon, and oxygen. [00116] In some embodiments, the additives can be added to the base material. In some embodiments, the additives are added while the base material is in the curing stage. In some embodiments, the additives are added via chemical vapor deposition after the base material has cured. In some embodiments, the additives are applied via a coating after the base material has cured. In some embodiments, the additives can be applied onto a surface or mold using 3D printing. In some embodiments, the radiation absorbing compositions of the present disclosure can be formed by 3D printing using the additives described herein.

[00117] In some embodiments, the additives are added to a mixture of the precursor and the one or more isotopes before adding the curing agent to the mixture. In some embodiments, the additives are added to a mixture of the precursor and the one or more isotopes before adding the curing agent to the mixture. In some embodiments, no additives are added. In some embodiments, after mixing in the curing agent to the precursor and the one or more isotopes, a base material is formed. In some embodiments, the additives can be added to the base material before, during, or after mixing.

[00118] Mixing [00119] The radiation absorbing compositions of the present disclosure can be prepared by various types of mixing. For instance, in some embodiments, the mixing is acoustic mixing. In some embodiments, acoustic mixing can be performed in an acoustic mixer, such as, for example, a RESONANTACOUSTIC® Mixer. In some embodiments, acoustic mixing allows for the application of low frequency, high- amplitude sound waves facilitating the movement of solids to induce mixing. In some embodiments, the mixing is high-speed shear mixing. In some embodiments, the mixing is a combination of high-speed shear mixing and acoustic mixing.

[00120] In some embodiments, the radiation absorbing compositions of the present disclosure can be prepared by various mixing mechanisms. For example, in some embodiments, the mixing mechanism can include, but is not limited to, paddle-blending, shaking, blending, solid suspension, turbines, close-clearance mixers, single-phase blending, high shear dispersers, static mixing, hand mixing, or combinations thereof.

[00121] In some embodiments, the mixing is performed such that the precursor, the neutron absorbing materials, and the curing agent are uniformly dispersed throughout the radiation absorbing composition. In some embodiments, the mixing is performed such that the additives are uniformly dispersed throughout the radiation absorbing composition. In some embodiments, better dispersion of smaller micro-particles or nanoparticles improves attenuation.

[00122] Applications and Advantages [00123] The present disclosure can have various advantages. For instance, in some embodiments, the radiation absorbing compositions of the present disclosure have at least the following valuable features: (i) the radiation absorbing compositions of the present disclosure have neutron absorbing capabilities at high temperatures; (ii) the radiation absorbing compositions of the present disclosure have low densities; (iii) the radiation absorbing compositions of the present disclosure have high neutron resistance; (iv) the radiation absorbing compositions of the present disclosure have high thermal conductivity; (v) the radiation absorbing compositions of the present disclosure have an effective neutron cross section; and (vi) the radiation absorbing compositions of the present disclosure exhibit low amounts of secondary gamma radiation. [00124] As such, the radiation absorbing compositions of the present disclosure can be utilized in various manners and for various purposes. For instance, in some embodiments, the radiation absorbing compositions can be used to absorb neutrons from an environment. In some embodiments, the radiation absorbing compositions are used to create or cover various containers for use in high radiation applications. For instance, in some embodiments, the containers are casks, such as, but not limited to, casks for storing or transporting radioactive material. In some embodiments, the containers may be nuclear waste casks. In some embodiments, the containers may be storage casks, transportation casks, dual-purpose transportation and storage casks, or combinations thereof.

[00125] In some embodiments, the casks can include an inner shell and an outer shell. In some embodiments, the space between the inner shell and the outer shell can include the radiation absorbing compositions. In some embodiments, the inner shell, the outer shell, or both the inner and outer shell include the radiation absorbing compositions of the present disclosure. In some embodiments, the casks can include a plurality of shells

[00126] In some embodiments, the present disclosure provides equipment or an equipment housing including the radiation absorbing compositions disclosed herein. In some embodiments, the equipment or equipment housing is applicable to the space or aerospace industry. For example, in some embodiments, the equipment may be used in airplanes, helicopters, or various aircraft. In some embodiments, the equipment or equipment housing is for the medical industry and can include, but is not limited to, shielding for radioactive testing and/or treatment to patients.

[00127] Additional Embodiments [00128] Reference will now be made to more specific embodiments of the present disclosure and experimental results that provide support for such embodiments. However, Applicants note that the disclosure below is for illustrative purposes only and is not intended to limit the scope of the claimed subject matter in any way.

[00129] Example 1. Computational and Experimental Comparison of Boron Carbide, Gadolinium Oxide, Samarium Oxide, and Graphene Platelets as Additives for a Neutron Shield [00130] This Example describes computational and experimental comparisons of boron carbide, gadolinium oxide, samarium oxide, and graphene platelets for use as additives for neutron absorbing and/or shielding.

[00131] Neutron shielding materials are used in aerospace, space, medical, and nuclear applications and there is a growing need of neutron shielding materials with increased temperature resistance. This is difficult to achieve with the current approach of using borated or lithium enriched polyethylene or polyamide composites because their hydrogen content limits temperature resistance to 200°C. To address this challenge, an epoxy-based neutron shielding material has been developed with a degradation temperature greater than 300°C and a macroscopic neutron cross section better than 5% borated polyethylene. By compensating for the low number of hydrogen atoms in this material, compared to polyethylene or polyamides, iron and additives with a high neutron absorption cross section such as gadolinium and samarium are incorporated to achieve the necessary attenuation properties. Graphene platelets were also added to evaluate graphene’s ability to mitigate secondary gamma from the neutron absorption by gadolinium and samarium. A computational study was first completed in Monte Carlo N- Particle (MCNP®) to benchmark existing materials and compare the loading of the potential additives. Neutron attenuation improved with increased loading of gadolinium oxide but not with increased loading of samarium oxide or boron carbide. Neutron irradiated samples show the flux attenuation for fast neutrons closely matches the computational model. The maximum fast neutron cross-sections for each absorber additive were 0.021 cm 1 in 9% B4C, 0.031 cm 1 with 27% Gd203, and 0.030 cm 1 with 3% Sm203 compared to only 0.023 cm 1 for 5% borated polyethylene. The sample containing 10% Gd203 and 2% B4C attenuated neutrons better than the borated polyethylene with a fast neutron cross section of 0.026 cm 1 and exhibited the lowest amount of secondary gamma photons of all the material combinations. [00132] Example 1.1. Introduction

[00133] Neutron shielding materials are important for protecting personnel and equipment in radioactive environments such as nuclear power plants, aerospace and space, or military activities. Due to the variety of constraints associated with the needs of different industries, many neutron shields have been developed ranging from hydrogen rich borated polymers to thermalize fast neutrons to metal foams or metal ceramic composites designed to work with a moderator, such as water, to absorb thermal neutrons. These materials often must perform in environments with high temperatures, corrosive conditions from radiation and acidic rainwater, and even weight restrictions, which pose challenging limitations on the design.

[00134] Advancements in the development of these materials are needed to reduce the mass or volume of shielding material necessary and enable the shield to operate at higher temperatures because existing shielding materials limit users’ options to either a low density or thermally stable product. For example, water, polyethylene and concrete are some of the most commonly used neutron shielding material in nuclear applications. Concrete is multi-purpose in shielding neutron radiation, shielding gamma radiation, and providing structural support or protection from impact. Concrete is stable at high temperatures but has a relatively high density between 2.3 and 6.4 g/cm3. Conversely, water has a density of 1.0 g/cm3 but begins to boil at l00°C and therefore requires cumbersome mechanical cooling systems. Similarly, borated polyethylene slows and absorbs neutrons well but has a temperature resistance less thanl80°C. The neutron attenuation abilities of water, concrete, and polyethylene have been leveraged with neutron absorbing additives such as boric acid, boron carbide, boron nanotubes, and ferro-boron but the density and/or temperature resistance are still limiting.

[00135] Boron Carbide, B4C, is a chemically and thermally stable ceramic. B4C is a common neutron absorbing additive and has been used in control rods and neutron absorption materials due to its chemical and thermal stability above 2,335°C, density of 2.3 g/cm3, B-10 content, and high absorption cross section for thermal neutrons. Since B4C is such a hard material, it is most readily available in powder form. Boron carbide has been used in neutron shielding materials ranging from ‘sandwich’ type boron carbide carbon fiber composites, ultra-high molecular weight polyethylene, B4C -PbO-Al(OH)3 epoxy nanocomposites, polyimide composites, and silicone rubber. Gadolinium oxide, Gd203, is an inorganic rare earth metal oxide with a density of 7.41 g/cm3 at room temperature and a melting point of 2420°C. Gadolinium has been used as a neutron absorbing additive to B4C-Al metals to improve the ductility and polymer composites to improve gamma shielding but it has not been as widely explored as boron containing additives. Similarly, Sm203, has a high density of 8.35 g/cm3 and melting point of 2,335°C but is even less common for neutron shielding than gadolinium oxide. It seems samarium oxide has only been incorporated into ultra-high molecular weight polyethylene, Portland Cement, and polyimide composites. [00136] Boron carbide, samarium oxide, and gadolinium oxide were selected as neutron absorption additives for comparison in this Example, and graphene platelets were incorporated to evaluate their effect on the secondary gamma photons produced from the neutron absorption reaction. Graphene has been used in polymer composites to improve the thermal conductivity or for shielding electromagnetic radiation. Natural samarium and natural gadolinium have higher thermal neutron absorption cross sections than boron- 10, but boron releases lower energy secondary gamma photons than the gadolinium. Boron carbide has historically been more readily available than samarium oxide and gadolinium oxide. Beyond material design constraints, an additional constraint is to maximize neutron stopping power while minimizing secondary gamma production. [00137] This Example presents an epoxy base with four types of additives used separately and in combination of varying weight percentages to evaluate options to minimize the total dose. The neat epoxy contains only 3% hydrogen, allowing it to resist higher temperatures than materials such as polyethylene which contain 14% hydrogen and make them effective at slowing fast neutrons. The neat epoxy also contains 22% iron, 28% silicon, 13% carbon and 34% oxygen. Table 1, shown below, compares the elements and isotopes used in the samples. Typically, the probability of neutron capture by boron, gadolinium, or samarium increases with decreasing neutron speed thus a neutron shield should also contain light elements, such as hydrogen, to therm alize the neutrons and increase their probability of capture. However, the approach in this Example is to rely on the iron in the epoxy to support neutron thermalization for absorption by the boron, samarium and gadolinium and to blend boron carbide and gadolinium to reduce the total photon dose in the gadolinium containing samples.

[00138] Borated polyethylene was used for neutron shielding benchmarking, in accordance with NASA’s (National Aeronautics and Space Administration) approach. 5% borated polyethylene has a high hydrogen content and enough boron to absorb the neutrons that have been thermalized by the hydrogen. Due to its effectiveness in neutron shielding, it is a most common polymer neutron shield and one of the goals in sample design was to achieve neutron attenuation properties comparable to 5% borated polyethylene.

Isotope/ 1 MeV 0.025 eV Total Neutron Notable Neutron

Element Total Total bound Absorption Absorption Reactions

Neutron Neutron scattering Reaction

Cross Cross cross 2200 m/s

Section Section section neutron

(barns) (barns)

From From From

Rinard Rinard NIST From NIST

Hydrogen 4.26 30.62 82.02 0.3326

B-10 2.68 3845 3.1 3835

Figure imgf000033_0001

B-l l 2.13 5.28 5.77 0.0055

Carbon 2.58 4.95 5.551 0.0035

Oxygen 8.22 3.87 4.232 0.00019

Natural 4.43 2.24 0.171 0.171

Silicon

Natural Iron 5.19 14.07 2.56 2.56

Natural 7.33 157

49,153 180 49,700 Gd+ nL|iemiai

Gadolinium 158Gd+y (7.88MeV) +ACK electrons (4.2keV)

Natural N/A N/A 39 5,922

Figure imgf000033_0002

Samarium 150Sm + 7.89 MeV

Table 1. A comparison of the neutron cross sections of the elements or isotopes included in the proposed neutron shielding composites.

[00139] Example 1.2. Computational Model in Monte Carlo N-Particle (MCNP®)

[00140] Prior to completing physical experiments, MCNP® was used to compare the neutron stopping power, gamma stopping power, and secondary gamma generation of potential shield formulations. An MCNP® code has three cards: a cell card to help define regions and the material associated with each cell, surface cards to complete the geometric configuration description, and data cards to call the types of particles, radiation sources, materials, tallies, cross section libraries and necessary level of detail for particle physics. The model has a neutron source in a void in the center of a spherical sample and a detector on the other side of the sample. The cross section of the sphere is schematically shown in FIG. 2, printed from MCNP®. The cell card, surface card, neutrons per second, mode, and tally definitions were held constant throughout material comparison. The source definition varied between 2.5E-8 MeV, thermal neutrons, and 2 MeV, fast neutrons. Runs were completed with a neutron detector to measure neutron attenuation of the shield and with a gamma detector to measure secondary gamma from neutron capture.

[00141] Example 1.3. Sample Fabrication

[00142] Samples were fabricated by incorporating varying amounts, 3 wt%, 9 wt%, and 27 wt% of the additives in a room temperature vulcanizing silicone rubber iron base (‘epoxy matrix’), including boron carbide nanopowder (from Sky Spring Nanomaterials, Inc., <50 nm), gadolinium oxide nanoparticles (from Sky Spring Nanomaterials, Inc., <100 nm), and samarium oxide nanoparticles (from Sky Spring Nanomaterials, Inc., -100 nm) and/or 1% Graphene Platelets (GP) (from Sky Spring Nanomaterials, Inc., 11-15 nm and electrical conductivity of 10 Siemens/meter parallel to the surface and 10 Siemens/meter perpendicular to the surface). 1 wt% Graphene Platelets were added to samples with 3 wt% and 9 wt% B4C, Gd203, and Sm203. The graphene platelet volume was too large to be incorporated with the base for samples containing 27 wt% of neutron absorbing additives. The benchmark sample of 5% borated polyethylene was obtained from Kings Plastic. The neat epoxy was modeled and fabricated in addition to the list of samples represented in Table 2, shown below.

B4C-based Additives Gd20 -based Additive Sm203-based Additive

3 wt% B4C 3 wt% Gd203 3 wt% Sm203

9 wt% B4C 9 wt% Gd203 9 wt% Sm203

27 wt% B4C 27 wt% Gd203 27 wt% Sm203

3 wt% B4C + 1 wt% GP 3 wt% Gd203 + 1 wt% GP 3 wt% Sm203 + 1 wt% GP 9 wt% B4C + 1 wt% GP 9 wt% Gd203 + 1 wt% GP 9 wt% Sm203 + lwt% GP

10 wt% Gd203 + 2 wt% B4C Table 2. A list of MCNP® modeled and fabricated additive combinations to epoxy matrix.

[00143] The nanoparticle size of the additives was selected based on a relationship between neutron absorber particle size and neutron attenuation at the same weight percentage of additive particle; attenuation improves as neutron absorbing particle size decreases due to the increase in aerial density of the neutron absorbing isotopes. As absorber weight percent increases neutron transmission tends to decrease, but with submicron nanoparticles the neutron transmission for 2% absorber loading is even less than neutron transmission for one-micron absorber particles at 5% loading, indicating submicron additive nanoparticles are preferred for neutron shielding. It was shown that by decreasing the particle size the macroscopic thermal neutron absorption cross- section increases because decreasing the particle size increases the collision probability between the incident thermal neutron and the shielding material, effectively improving the radiation shielding properties of material. Therefore, the smallest obtainable particle size of neutron absorbing additives was selected.

[00144] Neutron absorption is better optimized with more evenly distributed nanoparticles (again, due to the increase in neutron absorbing isotope aerial density), but evenly distributing nanoparticles in a highly viscous, rubber matrix is challenging due to the high surface energy of the nanoparticles in a polymer matrix. To address this challenge, an acoustic mixing process was selected as the method to incorporate the nanoparticles into the epoxy due to the high viscosity of the epoxy, the sub-micron size of the nanoparticles, and the high weight percent loading of the nanoparticles in the rubber. The nanoparticle distribution in the epoxy depends on particle size, particle shape, particle density, surface properties, powder cohesion, and variable bulk density. Mixing effectiveness varied for the different loadings of nanoparticles.

[00145] Example 1.4. Thermal Studies

[00146] Thermogravimetric analysis has been frequently used to study the decomposition temperature of silicone rubber and nanoparticle enriched silicone rubber materials and was used in this Example to compare the effect of the additives on the operating temperature of the epoxy. The percent weight loss of each sample was measured by heating the material to 400°C at a ramp of l0°C in an air environment. [00147] Example 1.5. Neutron Attenuation Experiments

[00148] Neutron attenuation was measured at the Rhode Island Nuclear Science Center using the R3 beam port containing neutrons in a half-inch, square, 2 MeV beamline from the 2 MW, light water cooled, pool type reactor (FIG. 3). The samples were placed in between the neutron beamline and a detector with paraffin moderating the fast neutrons for detection (FIG. 4). High- voltage is sent to the BF-3 detector from a power supply. It comes back through the pre-amp, the pre-amp goes to the amplifier, and the amplifier goes to the counter to measure relative intensity. The total flux was first calculated using a BF-3 detector with paraffin in front of it to slow the fast neutrons for detection by the B-10 in the BF-3 detector. The number of neutrons passing through each sample of known thickness (3”) was then determined using the same detector.

[00149] Example 1.6. Secondary Gamma Experiments

[00150] Secondary gamma was also modeled using MCNP® but a unique experiment was designed. A lead pig was lined with radiachromic film from Far West Technology, Inc. and samples of equal weight of each material were placed in a pig, one at a time. The pig was placed in front of the fast neutron beamline for twenty minutes, allowing sufficient time for the additives to absorb neutrons and produce secondary gamma. The radiachromic films were then cut (in a dark room) and placed in a radiachromic reader to measure the optical density of the film post irradiation. Five measurements from each radiachromic film were taken.

[00151] FIG. 5A illustrates a photo of the lead pig used for secondary gamma testing. The lead pig is 2 cm in diameter. FIG. 5B illustrates a schematic of the cross-sectional view of the sample in the lead pig lined with the radiachromic film.

[00152] Example 1.7. Computational Neutron and Photon Dose from MCNP®

[00153] From the MCNP® model it was observed that neutron dose rate in the epoxy decreased with increasing amounts of gadolinium oxide but not with increasing amounts of boron carbide or samarium oxide. The epoxy-based composites resulted in a lower secondary gamma dose rate, even with boron carbide, than borated polyethylene. This is most likely due to the iron and silicon in the epoxy absorbing the secondary gamma before it reached the detector. The addition of 3 or 9 wt% boron carbide in the epoxy reduced neutron dose slightly but neutron dose increased at 27 wt% boron carbide. This is due to the trade-off of reducing hydrogen content when increasing boron carbide content; the reduced amount of hydrogen in the samples hinders the material’s ability to slow the neutrons enough to increase the probability of neutron capture by the B-10 in the B4C. Gd203 was effective in absorbing neutrons with increased loading even though it delivers a higher secondary gamma dose. The Sm203 containing samples also caused higher secondary gamma than the B4C containing samples. The theoretical neutron dose rates are also shown; even though the secondary gamma was highest from the Gd203 the neutron dose was lower than the borated epoxy and Sm203 epoxy samples. To achieve a balance of neutron attenuation from the Gd203 and low secondary gamma from the B4C, a sample with 10 wt% Gd203 and 2 wt% boron carbide was modeled. Using gadolinium and boron will help reduce total dose because the boron can absorb the thermal neutrons without releasing high energy (MeV) gamma; it instead releases an alpha particle and lower energy gamma (keV). The combination of gadolinium and boron is effective because any neutrons not absorbed by the boron will have a higher probability of absorption by the gadolinium to minimize the total dose. The computational gadolinium oxide and boron carbide containing sample had a neutron dose equal to the sample with 9 wt% Gd203 but a photon dose rate of 4.59E-11 rem/hr instead of 5.44E-11 rem/hr.

[00154] FIG. 6 illustrates a comparison of the neutron dose (bar graph) and photon dose rates (line graph) on the outer surface of each material composition from the MCNP® simulation.

[00155] Example 1.8. Samples

[00156] Images of the cross section of samples with are shown in FIG. 7-FIG. 10. From these images, the even dispersion of the nanoparticles up to 27 wt% was feasible for Gd203 and Sm203 due to their high density compared to the base epoxy. The B4C nanoparticles were smaller than the Gd203 and Sm203 and darkened the epoxy with increasing weight percent loading. However, it was difficult to evenly distribute 27 wt % of boron carbide loading in the sample due to the high- volume ratio of boron carbide to polymer. Per FIG. 10, the boron carbide nanoparticles are visible in the 27 wt% sample due to the low density and high weight percentage of the boron carbide in the epoxy base. On the other hand, at 27 wt% loading, the Gd203 and Sm203 were significantly better incorporated than the boron carbide in the epoxy base.

[00157] Example 1.9. Temperature Resistance

[00158] The weight loss curve for all of the samples was consistent with the curve of the epoxy base, except for the 27 wt% B4C loaded sample. B4C is significantly less dense than gadolinium oxide and samarium oxide making it difficult to evenly distribute 27 wt% B4C in the epoxy within the working time. The uneven distribution of B4C in the epoxy, highlighted in FIG. 10, affected the temperature resistance of the epoxy and caused a melting point depression in the epoxy. However, the other samples - like the neat epoxy - did not start to decompose until >300°C showing that adding more boron carbide to the epoxy significantly reduces operating temperature. A commercially available neutron shielding material, SWX-227, is a borated silicone rubber that was measured for benchmarking for thermal studies.

[00159] FIG. 11A illustrates TGA curve for all samples matched the degradation of the epoxy curve (shown) except the 27 wt% B4C sample due to the challenges in evenly distributing the nanoparticles. FIG. 11B illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% B4C in epoxy. FIG. 11C illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% Gd203 in epoxy. FIG. 11D illustrates weight loss curve for 3 wt%, 9 wt% and 27 wt% Sm203 in epoxy.

[00160] Example 1.10. Neutron Attenuation

[00161] Neutron attenuation results for different neutron absorbing additives in the epoxy are presented in FIG. 12. A 5% borated polyethylene sample was used for benchmarking. It is shown that the even with loading the maximum amount of boron carbide in the epoxy, fast neutron attenuation is still not as good as the borated polyethylene. In contrast, the epoxy composites containing gadolinium or samarium compensated for the low amount of hydrogen and were even more effective than the borated polyethylene at attenuating neutrons. Furthermore, using gadolinium oxide and samarium oxide also avoids the thermal stability problem revealed in FIG. 10 for high concentrations of B4C additives. [00162] Neutron attenuation did not improve with increasing boron carbide loading because the total amount of hydrogen in the sample decreased and the neutrons were not slow enough to increase the probability of capture by the B-10 in the B4C. With gadolinium, neutron attenuation improved with increasing gadolinium oxide loading because gadolinium has a high neutron cross section for thermal and fast neutrons and the decrease in hydrogen and iron content is not as necessary for the gadolinium containing samples as it is for the boron containing samples. Samples with as little as 3 wt% gadolinium oxide demonstrated better neutron attenuation than the borated polyethylene sample. Similarly, samples with even 3 wt% samarium oxide demonstrated better neutron attenuation than the borated polyethylene sample. The samarium oxide has a lower probability of neutron capture for fast neutrons than gadolinium oxide thus neutron attenuation was better for samples with a lower samarium oxide loading. Since the MCNP® results indicated high secondary gamma from the gadolinium, one sample was made combining 10 wt% gadolinium oxide and 2 wt% boron carbide to synergistically harness the effectiveness of gadolinium oxide for neutron absorption and the boron carbide for secondary gamma reduction. As shown in FIG. 12 the neutron attenuation of this sample is better than 9% Gd203 as well as all the B4C-doped samples.

[00163] The ratio of the neutron flux with shielding to that without shielding, Flo, can be used to calculate the macroscopic neutron absorption coefficient from Beer-Lambert, Equation 1 below, where x is the thickness of the sample.

Equation 1 å(E) = - iln( )

x i 0

[00164] Table 3, shown below, illustrates experiment macroscopic fast neutron cross-sections.

Sample

Figure imgf000039_0001

5% B Poly 0.023 ± 0.005

Epoxy Neat 0.020 ± 0.002

3% B4C in epoxy 0.020 ± 0.002

9% B4C in epoxy 0.021 ±0.003 27% B4C in epoxy 0.019 ± 0.004

3% Gd203 in epoxy 0.023 ± 0.003

9% Gd203 in epoxy 0.024 ± 0.005

27% Gd203 in epoxy 0.031 ± 0.001

3% Sm203 in epoxy 0.030 ± 0.005

9% Sm203 in epoxy 0.028 ± 0.003

27% Sm203 in epoxy 0.026 ± 0.005

10% Gd203 and 2% B4C in epoxy 0.027 ± 0.002

Table 3. Experimental macroscopic fast neutron cross-sections.

[00165] Compared to the computational model trends, the experimental results have consistent trends among the samples with different additives. An interesting experimental result is that all the samples incorporating Gd203 and Sm203 achieved better neutron attenuation than the borated polyethylene sample. As discussed earlier, particle size may affect the attenuation in the model compared to the experiments which may further explain the consistency between samples with the same type of additive. The homogeneous nanoparticle assumption in MCNP® may under estimate the macroscopic cross section of the samples. However, considering the experimental deviations, the results are within the computational model. [00166] When graphene platelets were added to the samples containing 3 wt% and 9 wt% neutron absorber, with the aim of reducing secondary gamma, the samples were measured for neutron attenuation to ensure the graphene platelets did not hinder neutron shielding performance. As previously described, incorporating graphene platelets and neutron absorbing additives made it difficult to evenly disperse the neutron absorbing nanoparticles and graphene platelets in the epoxy base. The nanoparticle agglomeration that accompanied the graphene platelets may be the cause for the slightly reduced neutron attenuation performance in most of the samples containing graphene platelets. Only the 3% gadolinium oxide sample improved with the graphene platelets but even that sample was within the standard deviation of the 3% gadolinium oxide sample without graphene platelets. [00167] FIG. 13 shows relative neutron transmission of a 2 MeV beamline through 3” thick samples containing neutron absorbing additives and graphene platelets.

[00168] Example 1.11. Experimental Secondary Gamma

[00169] FIG. 14 indicates photon dose on the film from secondary gamma and the ability of the material to shield gamma from the reactor. The lead pig lined with radiachromic film‘empty’ absorbed some gamma from the reactor. When materials were placed in a lead pig with radiachromic film, regardless of the material combination, gamma dose to the radiachromic film was reduced. Photon dose to the radiachromic film did not necessarily increase with increasing neutron absorbing additive, unlike the MCNP® model prediction shown in FIG. 6. This is because the computational model only detected secondary gamma. When gamma and neutrons from the reactor are incident on the neutron shield, all the samples are effective at reducing total photon dose regardless of neutron absorbing additive loading or density of the total material. However, as indicated by the computational results in FIG. 6, the combination of 10 wt% gadolinium oxide and 2 wt% boron carbide in the epoxy significantly reduces the photon dose over the 9 wt% gadolinium oxide epoxy and even more so than the 9 wt% boron carbide epoxy sample. The purpose of the addition of graphene platelets (GP) to samples contain 3 wt% and 9 wt% of neutron absorber was to reduce the total gamma dose. Per FIG. 14, a reduction in secondary gamma due to the addition of graphene platelets was not demonstrated in this Example. Overall, consistent with expectations, the combination of 10 wt% gadolinium oxide and 2 wt% boron carbide in the epoxy achieved the best gamma photon attenuation and high neutron attenuation by balancing neutron shielding and secondary gamma photon reduction.

[00170] Example 1.12. Conclusion

[00171] Boron carbide, gadolinium oxide and samarium oxide were compared for effectiveness as neutron absorbing additives in a neutron shield. A key finding is that an epoxy containing less hydrogen than polyethylene can be used for more effective neutron attenuation than borated polyethylene at higher temperatures if gadolinium or samarium is used for neutron absorption. Another key finding is that boron carbide can reduce secondary gamma in samples containing gadolinium oxide and can reduce the secondary gamma to levels below those in borated polyethylene.

[00172] The samples with 9% and 27% Gd203 had high neutron stopping power but a higher associated photon flux and the sample with 27% Gd203 and 3% Sm203 had the highest neutron stopping power. At 2 MeV all the samples fabricated with neutron absorbing additives reduced the photon dose than borated polyethylene samples. The sample with 10% Gd203 and 2% B4C had the best neutron stopping power with the lowest photon dose, lower than the 5% borated polyethylene material. This is important for the design of future neutron shielding materials because combining boron carbide and gadolinium oxide in a neutron shield addresses the secondary gamma disadvantage of using gadolinium oxide. The samarium oxide samples also had better neutron stopping power and a lower photon dose than the 5% borated polyethylene material. These results suggest samarium oxide may be a better additive than the gadolinium oxide because neutron stopping power comparable to borated polyethylene can be achieved and the total photon dose can also be reduced. Graphene platelets did not provide a noteworthy improvement in the neutron attenuation of any samples. The graphene platelets instead made fabrication more challenging due to their high surface area and the volume of the platelets that had to be evenly incorporated into the samples.

[00173] Advantages related to mechanical properties, thermal conductivity, or total neutron/gamma resistance in a neutron shield by using various sizes of graphene platelets, carbon nanofibers, carbon nanotubes, or buckeye balls additives is envisioned. Additionally, since the assumed homogeneous distribution of neutron absorbing isotopes in MCNP® may have underestimated the macroscopic cross section of the material, it is envisioned that optimum particle sizes, considering attenuation and fabrication, can be identified by evaluation of the effect of nanoparticle size on attenuation. Furthermore, neutron absorbing compositions with samarium, gadolinium, and boron compounds as neutron shielding additives in different types of polymeric and metal bases are further envisioned.

[00174] Without further elaboration, it is believed that one skilled in the art can, using the description herein, utilize the present disclosure to its fullest extent. The embodiments described herein are to be construed as illustrative and not as constraining the remainder of the disclosure in any way whatsoever. While the embodiments have been shown and described, many variations and modifications thereof can be made by one skilled in the art without departing from the spirit and teachings of the invention. Accordingly, the scope of protection is not limited by the description set out above, but is only limited by the claims, including all equivalents of the subject matter of the claims. The disclosures of all patents, patent applications and publications cited herein are hereby incorporated herein by reference, to the extent that they provide procedural or other details consistent with and supplementary to those set forth herein.

Claims

WHAT IS CLAIMED IS:
1. A radiation absorbing composition, the composition comprising:
a base material; and
one or more isotopes associated with the base material,
wherein the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns.
2. The composition of claim 1, wherein the one or more isotopes comprise two or more different isotopes of a same element, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
3. The composition of claim 1, wherein the one or more isotopes comprise two or more different isotopes of different elements, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
4. The composition of claim 1, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, and a ceramic.
5. The composition of claim 1, wherein the one or more isotopes mitigate negative effects of neutron absorption, wherein the negative effects are selected from the group consisting of secondary gamma production by the composition after neutron absorption, alpha production by the composition after neutron absorption, secondary x-ray production by the composition after neutron absorption, and combinations thereof.
6. The composition of claim 1, wherein the one or more isotopes are part of one or more neutron absorbing compounds.
7. The composition of claim 6, wherein the one or more neutron absorbing compounds are selected from the group consisting of boron-containing materials, gadolinium- containing materials, samarium-containing materials, cadmium-containing materials, molybdenum-containing materials, hafnium-containing materials, titanium-containing materials, dysprosium-containing materials, iron-containing materials, lithium-containing materials, ytterbium-containing materials, zinc-containing materials, or combinations thereof.
8. The composition of claim 6, wherein the one or more neutron absorbing compounds comprise at least one of boron carbide, boric acid, boron nitride, boron oxide, gadolinium oxide, gadolinium acetate, samarium oxide, cadmium oxide, molybdenum boride, hafnium diboride, titanium diboride, dysprosium titanate, lithium, gadolinium titanate, iron oxide, or combinations thereof.
9. The composition of claim 6, wherein the one or more neutron absorbing compounds comprise a gadolinium-containing material and a boron containing material.
10. The composition of claim 6, wherein the one or more neutron absorbing compounds comprise gadolinium oxide and boron carbide.
11. The composition of claim 6, wherein the one or more neutron absorbing compounds are in the form of nanoparticles.
12. The composition of claim 1, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of boron, gadolinium, cadmium, samarium, lithium, hafnium, cobalt, titanium, dysprosium, erbium, europium, molybdenum, ytterbium, zinc, and iron.
13. The composition of claim 1, wherein the one or more isotopes are natural or enriched isotopes selected from the group consisting of lithium-6 (6Li), lithium-7 (7Li), boron- 10 (10B), boron-l l (nB), gadolinium- 157 (157Gd), gadolinium- 152 (152Gd), gadolinium- 154 (154Gd), gadolinium- 155 (155Gd), gadolinium- 156 (156Gd), gadolinium- 158 (158Gd), gadolinium- 160 (160Gd), gadolinium- 185 (185Gd), samarium-l5l (151Sm), samarium-l49 (149Sm), samarium-l44 (144Sm), samarium-l50 (150Sm), samarium-l52 (152Sm), samarium-l54 (154Sm), cadmium-H3 (113Cd), cadmium-H6 (116Cd), cadmium-l06 (106Cd), cadmium-l08 (108Cd), cadmium-H4 (114Cd), cadmium-l lO (110Cd), cadmium-l l l (mCd), cadmium-H2 (112Cd), cadmium-l09
(109Cd), cadmium-H5 (115Cd), and cadmium-H7 (117Cd).
14. The composition of claim 1, wherein the one or more isotopes are uniformly dispersed throughout the base material.
15. The composition of claim 1, wherein the composition has a density at or below 2 g/cm3.
16. The composition of claim 6, wherein the composition has a hydrogen content of less than about 10% by combined weight of the base material and the neutron absorbing compounds.
17. A method of absorbing neutrons from an environment, wherein the method comprises:
associating the environment with a composition, wherein the composition comprises: a base material, and
one or more isotopes associated with the base material,
wherein the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns; and
wherein the associating results in the absorption of the neutrons from the environment onto the composition.
18. The method of claim 17, wherein the one or more isotopes comprise two or more different isotopes of a same element, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
19. The method of claim 17, wherein the one or more isotopes comprise two or more different isotopes of different elements, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
20. The method of claim 17, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, and a ceramic.
21. The method of claim 17, wherein the associating comprises flowing the environment through the composition.
22. The method of claim 17, wherein the environment is an atmosphere of a nuclear power plant.
23. The method of claim 17, wherein the one or more isotopes mitigate negative effects of neutron absorption, wherein the negative effects are selected from the group consisting of secondary gamma production by the composition after neutron absorption, alpha production by the composition after neutron absorption, secondary x-ray production by the composition after neutron absorption, and combinations thereof.
24. The method of claim 17, wherein the one or more isotopes are part of one or more neutron absorbing compounds.
25. The method of claim 24, wherein the one or more neutron absorbing compounds are selected from the group consisting of boron-containing materials, gadolinium-containing materials, samarium-containing materials, cadmium-containing materials, molybdenum- containing materials, hafnium-containing materials, titanium-containing materials, dysprosium- containing materials, iron-containing materials, lithium-containing materials, ytterbium- containing materials, zinc-containing materials, or combinations thereof.
26. The method of claim 24, wherein the one or more neutron absorbing compounds comprise at least one of boron carbide, boric acid, boron nitride, boron oxide, gadolinium oxide, gadolinium acetate, samarium oxide, cadmium oxide, molybdenum boride, hafnium diboride, titanium diboride, dysprosium titanate, lithium, gadolinium titanate, iron oxide, or combinations thereof.
27. The method of claim 24, wherein the one or more neutron absorbing compounds comprise a gadolinium-containing material and a boron containing material.
28. The method of claim 24, wherein the one or more neutron absorbing compounds comprise gadolinium oxide and boron carbide.
29. The method of claim 24, wherein the one or more neutron absorbing compounds are in the form of nanoparticles.
30. The method of claim 17, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of boron, gadolinium, cadmium, samarium, lithium, hafnium, cobalt, titanium, dysprosium, erbium, europium, molybdenum, ytterbium, zinc, and iron.
31. The method of claim 17, wherein the one or more isotopes are natural or enriched isotopes selected from the group consisting of lithium-6 (6Li), lithium-7 (7Li), boron- 10 (10B), boron-l l (nB), gadolinium- 157 (157Gd), gadolinium- 152 (152Gd), gadolinium- 154 (154Gd), gadolinium- 155 (155Gd), gadolinium- 156 (156Gd), gadolinium- 158 (158Gd), gadolinium- 160 (160Gd), gadolinium- 185 (185Gd), samarium-l5l (151Sm), samarium-l49 (149Sm), samarium-l44
(144Sm), samarium-l50 (150Sm), samarium-l52 (152Sm), samarium-l54 (154Sm), cadmium-H3 (113Cd), cadmium-H6 (116Cd), cadmium-l06 (106Cd), cadmium-l08 (108Cd), cadmium-H4 (114Cd), cadmium-l lO (110Cd), cadmium-l l l (mCd), cadmium-H2 (112Cd), cadmium-l09 (109Cd), cadmium-H5 (115Cd), and cadmium-H7 (117Cd).
32. The method of claim 17, wherein the one or more isotopes are uniformly dispersed throughout the base material.
33. The method of claim 17, wherein the composition has a density at or below 2 g/cm .
34. The method of claim 24, wherein the composition has a hydrogen content of less than about 10% by combined weight of the base material and the neutron absorbing compounds.
35. A method of preparing a radiation absorbing composition, the method
comprising:
mixing a precursor material with one or more isotopes and a curing agent; and wherein the mixing results in the formation of a base material from the precursor material,
wherein the base material becomes associated with the one or more isotopes, and wherein the one or more isotopes have an individual or combined thermal neutron cross section of more than 50 barns.
36. The method of claim 35, wherein the one or more isotopes comprise two or more different isotopes of a same element, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
37. The method of claim 35, wherein the one or more isotopes comprise two or more different isotopes of different elements, where a combined thermal neutron cross section of the two or more isotopes is more than 50 barns.
38. The method of claim 35, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of a metal, a metalloid, a transition metal, a post-transition metal, a lanthanide, an oxide, and a ceramic.
39. The method of claim 35, wherein the curing agent is selected from the group consisting of an amine-type hardener, an acid anhydrate-type hardener, an imidazole-type hardening promoter, water, or combinations thereof.
40. The method of claim 35, wherein the curing agent is an amine-type hardener selected from the group consisting of an aromatic amine, an alicyclic amine, a polyamide amine, or combinations thereof.
41. The method of claim 35, wherein the one or more isotopes are part of one or more neutron absorbing compounds.
42. The method of claim 41, wherein the one or more neutron absorbing compounds are selected from the group consisting of boron-containing materials, gadolinium-containing materials, samarium-containing materials, cadmium-containing materials, molybdenum- containing materials, hafnium-containing materials, titanium-containing materials, dysprosium- containing materials, iron-containing materials, lithium-containing materials, ytterbium- containing materials, zinc-containing materials, or combinations thereof.
43. The method of claim 41, wherein the one or more neutron absorbing compounds comprise at least one of boron carbide, boric acid, boron nitride, boron oxide, gadolinium oxide, gadolinium acetate, samarium oxide, cadmium oxide, molybdenum boride, hafnium diboride, titanium diboride, dysprosium titanate, lithium, gadolinium titanate, iron oxide, or combinations thereof.
44. The method of claim 41, wherein the one or more neutron absorbing compounds comprise a gadolinium-containing material and a boron containing material.
45. The method of claim 41, wherein the one or more neutron absorbing compounds comprise gadolinium oxide and boron carbide.
46. The method of claim 41, wherein the one or more neutron absorbing compounds are in the form of nanoparticles.
47. The method of claim 35, wherein the one or more isotopes are derived from one or more elements selected from the group consisting of boron, gadolinium, cadmium, samarium, lithium, hafnium, cobalt, titanium, dysprosium, erbium, europium, molybdenum, ytterbium, zinc, and iron.
48. The method of claim 35, wherein the one or more isotopes are natural or enriched isotopes selected from the group consisting of lithium-6 (6Li), lithium-7 (7Li), boron-lO (10B), boron-l l (nB), gadolinium- 157 (157Gd), gadolinium- 152 (152Gd), gadolinium- 154 (154Gd), gadolinium- 155 (155Gd), gadolinium- 156 (156Gd), gadolinium- 158 (158Gd), gadolinium- 160 (160Gd), gadolinium- 185 (185Gd), samarium-l5l (151Sm), samarium-l49 (149Sm), samarium-l44 (144Sm), samarium-l50 (150Sm), samarium-l52 (152Sm), samarium-l54 (154Sm), cadmium-H3 (113Cd), cadmium-H6 (116Cd), cadmium-l06 (106Cd), cadmium-l08 (108Cd), cadmium-H4
(114Cd), cadmium-l lO (110Cd), cadmium-l l l (mCd), cadmium-H2 (112Cd), cadmium-l09 (109Cd), cadmium-H5 (115Cd), and cadmium-H7 (117Cd).
49. The method of claim 35, wherein the one or more isotopes are uniformly dispersed throughout the base material.
50. The method of claim 35, wherein the composition has a density at or below 2 g/cm .
51. The method of claim 41, wherein the composition has a hydrogen content of less than about 10% by combined weight of the base material and the neutron absorbing compounds.
52. The method of claim 35, wherein a component selected from the group consisting of the precursor material, the one or more isotopes, the curing agent, one or more neutron absorbing compositions, or combinations thereof is applied onto a surface or mold using three-dimensional (3D) printing.
PCT/US2019/027492 2018-04-13 2019-04-15 Neutron shielding and absorption materials WO2019200386A1 (en)

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