Nuclear Structural Aging

Wonjin Yun
March 12, 2016

Submitted as coursework for PH241, Stanford University, Winter 2016


Fig. 1: Schematic of typical boiling water reactor. (Source: Wikimedia Commons)

The demanding environments of an operating nuclear reactor may impact the ability of a broad range of materials to perform their intended function over extended service periods. Identifying materials and components where degradation may occur is an important aspect of safe and secure operation of a nuclear power plant. According to 2015 Nuclear Power Reactors in the World from IAEA [1], the global average operating time of a nuclear power plant is reached at 28 years. In order for many nuclear power plants to be able to operate safely beyond their initial design life, each utility has to demonstrate that high levels of safety and apply the advanced ageing management techniques. Thus it is essential to understand the mechanisms of radiation damage to the system structure and components and possible mitigation approaches.

Nuclear Power Plant Structure

As shown Fig. 1, commercial nuclear power plants (NPPs) in the United States contain concrete structures whose performance and function are necessary for the protection and safety of plant operating personnel, the general public, and the environment. Typical safety-related concrete structures contained in boiling-water reactor (BWR) plants are generally grouped into four categories: primary containments, containment internal structures, secondary containments/reactor buildings, and other structures. [2] These safety-related concrete and other structures must be capable of maintaining structural integrity for the operating life of the plant. Nuclear safety-related concrete structures are composed of several constituents that, in concert, perform multiple functions. Primarily, these constituents include the following material systems: concrete, conventional steel reinforcement, prestressing steel, steel liner plate, and embedment steel.

Primary Containment of Boiling-water Reactors (BWR)

Primary containment is primarily required to provide an "essentially" leak-tight barrier against the uncontrolled release of radioactivity to the environment. BWR plants utilize either reinforced or pre-stressed concrete primary containments. [2] A steel liner attached to the containment inside surface by studs or by structural steel members provides leak tightness of each of these containments. Exposed surfaces of the carbon steel liner are typically painted to protect against corrosion and to facilitate decontamination. A portion of the liner toward the bottom of the containment and over the basement is typically embedded in concrete to protect it from damage, abrasion, etc. due to corrosive fluids and impact. A seal to prevent the ingress of fluids is provided at the interface around the circumference of the containment where the vertical portion of the liner becomes embedded in the concrete.


Irradiation by charged particles such as electrons and protons can lead directly to a gain or a loss of charge of an atom, which is called direct ionizing radiation (ionization is the process of an atom gaining or losing electrons). Irradiation by noncharged particles such as neutrons and photons does not lead directly to a change in charge of an atom, and thus is called indirect ionizing radiation. Although uncharged particles do not cause direct ionization of atoms, they do release charged particles (electrons or protons) as a consequence of their interaction with matter and will subsequently cause direct ionization of atoms. Therefore, both direct and indirect ionizing radiation can cause ionization. In direct ionizing irradiation, the charged particles interact strongly with the materials in a radiation shield and are attenuated by the shield. In indirect ionizing irradiation, noncharged particles such as neutrons and gamma photons interact less strongly with shielding materials. [3]

Impact of Irradiation on Concrete

Gamma rays can reduce the water content in a shielding material such as concrete by two mechanisms. The gamma rays can decompose water through a process called radiolysis which converts the water to hydrogen, oxygen, and hydrogen peroxide. [3] This can affect concrete's creep and shrinkage behavior to a limited extent and also result in evolution of gas. Water can also be driven from concrete by evaporation caused by the heat generated by gamma-ray irradiation.

Neutron's interaction with nuclei of atoms may change the lattice spacing within the material after the collision of neutrons with nuclei. [3] Therefore, neutrons have a more significant effect on dense and well-crystallized materials than on randomly structured materials with high porosity. In concrete, aggregates (coarse and fine aggregates) are in a crystallized phase, and cement paste is an amorphous phase, and thus, neutron radiation causes more distortion and damage to the internal structure of aggregates than to the structure of cement paste. Therefore, prolonged exposure of concrete to irradiation can result in decreases in tensile and compressive strengths and modulus of elasticity. [2]

Impact of Irradiation on Steel

Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation or RIS) of these steels and degrades their fracture properties. Chopra has addressed the irradiation-induced materials issues such as irradiation-assisted stress corrosion cracking (IASCC) and irradiation-induced stress relaxation. [4] He describes the mechanism of producing damage by neutron irradiation of austenitic SSs. [4] Neutron irradiation displacing atoms from their lattice position, which creates point defects such as vacancies and interstitials. The defects rearrange into more stable configurations such as dislocation loops, network dislocations, precipitates, and cavities (or voids). Due to complexity of the irradiation-induced materials change for austenitic steels (SSs), Chopra investigated and identified the key material and environmental parameters that influence these processes and to determine their effects. [4] For instance, it was observed that the yield strength of irradiated SSs can increase up to five times that of the nonirradiated material after a neutron dose of about 5 dpa. Loss of preload for bolting and springs due to irradiation-induced stress relaxation is another aging degradation effect that needs to be addressed to assure the functional integrity of the reactor core internal components. Stress relaxation represents plastic deformation that occurs under constant strain below the yield point of the material. The literature pointed out that neutron-irradiation-enhanced creep at PWR temperatures can greatly increase the plastic strain by increasing both the transient creep and steady-state creep rates. [3]

© Wonjin Yun. 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] "Nuclear Power Reactors in the World," International Atomic Energy Agency, IAEA-RDS-2/35, May 2015.

[2] D. J. Naus, C. B. Oland, and B. R. Ellingwood, "Report on Aging of Nuclear Power Plant Reinforced Concrete Structures," U.S. Nuclear Regulatory Commission, NUREG/CR-6424, March 1996.

[3] K. William, Y. Xi, and D. Naus, "A Review of the Effect of Radiation on Microstructure and Properties of Concretes Used in Nuclear Power Plants," U.S. Nuclear Regulatory Commission, NUREG/CR-7171, November 2013.

[4] D. K. Chopra, "Degradation of LWR Core Internal Materials due to Neutron Irradiation," U.S. Nuclear Regulatory Commission, NUREG/CR-7027, December 2010.