Ceramic Radiation Shields

Andrew Lange
March 18, 2011

Submitted as coursework for Physics 241, Stanford University, Winter 2011

Fig. 1: C/SiC ceramic composite body flaps of NASA's experimental spacecraft X-38. Source: Wikimedia Commons

Nuclear reactor cores are very violent places. Huge numbers of high energy neutrons bombard their neighbors causing an eruption of gammas, betas, alphas, and more neutrons. It is important to stop the gamma rays and fast neutrons before they interact with biological structures for which ionization can lead to sickness and death. Historically, a layer of heavy metal such as lead and a few meters of concrete were used to shield radiation, however they are subject to damage after long periods of exposure and intense heat. To find ideal material candidates for radiation shielding, one must consider a number of properties: heat resistance, heat diffusivity, how well the material maintains it's structure after long periods of irradiation, ability to capture high speed and thermal neutrons, gamma ray attenuation coefficients, cost, ability to mold, adhere, etc.. This is a long list of properties for one material. Therefore, it is common to design multiple layers of shielding structures. In the last decade or so, ceramic composite materials have increasingly been proposed as a desirable reactor material (both inside and outside the core). These materials are easily tunable, strong, currently produced for several industrial applications, and extremely insensitive to high temperature conditions. There is also considerable irradiation data available because of electronic substrate applications in nuclear experiments. The following paragraphs provide some theoretical and experimental support for the use of composite materials in the construction of nuclear reactors. A brief discussion is also included outlining potential limitations of these materials.

The escape of potentially harmful, high energy, neutrons is limited by the reactor moderator (e.g. water, graphite). Because of it's high scattering cross-section of approximately 11 barns, Iron has been used as an additional attenuator of reactor core neutron flux. [1] This could potentially be used in conjunction with a ceramic shield or the composite could be doped with Iron atoms. However, a thick enough carbon based ceramic shield (C thermal cross-section - 5.551 barns) would sufficiently limit the transmission of neutrons. [2] In fact this may lead to problems if the carbon "burns up" faster than other elements in the cermaic (e.g. silicon).

Gamma ray absorption is therefore of primary concern when evaluating radiation shield materials. There are three physical processes that dominate gamma ray attenuation - compton scattering, pair production, and the photoelectric effect. [3] The total probability that a gamma ray will interact with a given material is just the sum of the individual probabilities that these three interactions occur: Pcs, Ppp, Ppe.

Pint (cm-1) = Pcs + Ppp + Ppe

Pint is the linear attenuation coefficient and gives the transmitted gamma radiation intensity through a material of thickness &Delta x by the Beer-Lambert equation:

I = I0e-Pint&Delta x

Gamma ray attenuation is energy dependent and will depend on which of the three processes dominates(s). Ppe ∝ Z5, Pcs ∝ Z, and Ppp ∝ Z2, where Z is the atomic number. [4] For lower energy gamma rays (less than 0.1 MeV), the photoelectric effect and pair production will play a significant role and heavy materials will attenuate substantially better than light materials. However, at energies above 5 MeV, compton scattering dominates and material weight is less significant (note: gamma ray energy bursts in reactors near 7 MeV). The table below gives an idea of common numbers for these coefficients. Note that to first order, a carbon based shield with density about that of graphite has to be about 1.5 times as thick as a lead shield to attenuate gammas equally.

Lead Iron Aluminum Graphite Water
Theory Pint (cm-1) 0.102 0.0718 0.0763 0.079 0.085
Experiment Pint (cm-1) 0.092 ± 0.001 0.0718 ± 0.001 0.070 ± 0.002 0.0720 ± 0.0002 0.0680 ± 0.0001
Table 1: Measured and predicted attenuation coefficients determined by measuring only photoelectric peaks. The measurements were taken using Cs137, 0.661MeV source. [4]

It is important that shielding materials do not become brittle and retain their compression strength after long exposure to violent core conditions. This was the impetus for using materials other than the historically popular cement and metal mixes. In fact, cured concrete used in early reactors operated at 50% of it's nominal compression strength at temperatures around 450°C. [1] Many fiber reinforced composites retain their mechanical properties (fracture toughness, elastic modulus) after being thermally aged at temperatures up to 1000°C. Nextel 720 fiber Alumino-silicate and Ti3SiC2/SiC are such examples. [5,6]

Porous Matrix Ceramic Composite SiC composite Concrete
Flexural Strength at 1000° C (MPa) No Data &asymp 500 [5] 2-5
Tensile Strength at 1000° C (MPa) &asymp 130 [6] No Data 40 - 80
Table 1: Flexural and tensile strengths of ceramic composite materials and broad concrete paramters. The figures for concrete were amalgamated for a range of mixtures.

It is obvious from the table above that ceramic composites are much stronger than traditional concretes even after about 1000 hours of exposure to extremely high temperatures. This is especially true for the flexural, or bending strength of the materials which is of great importance in nuclear reactor design because of their spherical or cylindrical geometries.

We see that ceramic materials, particularly carbon based composites, have many desirable structural properties for intense environments like those in nuclear reactors. It appears as if they should moderate neutrons and gamma radiation moderately well. However, further research into specific structures, their burn-up rates, and their effective cross-sections is necessary. Regardless of the results, they should be considered for intermediary layers in reactors designs to help mediate thermal fluxes and thermal expansion.

© Andrew Lange. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.


[1] J. F. Lamond and J. H. Pielert Significance of Tests and Properties of Concrete and Concrete Making Material (ASTM International, 2006).

[2] V. F. Sears, "Neutron Scattering Lengths and Cross Sections", Neutron News 3, 29 (1992).

[3] M. Jalali and A. Mohammadi, "Gamma Ray Attenuation Coefficient Measurement for Neutron-Absorbent Materials," Radiation Physics and Chemistry 77, 523 (2008).

[4] E. Bashandy, "Experimental Determination of the Absorption Coefficients of Gamma Rays through Different Barriers", Int. J. Appl. Radiation and Isotopes 13, 173 (1962).

[5] T. Yano et al., "Synthesis and High Temperature Mechanical Properties of Ti3SiC2/SiC Composite", J. Mat. Sci. 30, 3097 (1995).

[6] E. A. V. Carelli et al., "Effects of Thermal Aging on the Mechanical Properties of a Porous-Matrix Ceramic Composite," J. Am. Ceram. Soc. 85, 595 (2002).