Neutrons with sufficient energy can disrupt the atomic arrangement or crystalline structure of materials. The influence of structural damage is most significant for metals because of their relative immunity to damage by ionizing radiation. Pressurized-water reactors operate with a higher rate of neutron impacts and their vessels therefore tend to experience a greater degree of embrittlement than boiling-water reactor vessels. Many pressurized-water reactors design their cores to reduce the number of neutrons hitting the vessel wall. This slows the vessel’s embrittlement. The NRC’s regulations address embrittlement in 10 CFR Part 50, Appendix G, “Fracture Toughness Requirements” and Appendix H, “Reactor Vessel Material Surveillance Program Requirements.” Since the reactor pressure vessel is considered irreplaceable, neutron irradiation embrittlement of pressure vessel steels is a key issue in the long term assessment of structural integrity for life attainment and extension programmes.
Radiation damage is produced when neutrons of sufficient energy displace atoms (especially in steels at operating temperatures 260 – 300°C) that result in displacement cascades which produce large numbers of defects, both vacancies and interstitials. Although the inside surface of the RPV is exposed to neutrons of varying energies, the higher energy neutrons, those above about 0.5 MeV, produce the bulk of the damage. In order to minimize such material degradation type and structure of the steel must be appropriately selected. Today it is known that the susceptibility of reactor pressure vessel steels is strongly affected (negatively) by the presence of copper, nickel and phosphorus.
As was written, the distinction between brittleness and ductility isn’t readily apparent, especially because both ductility and brittle behavior are dependent not only on the material in question but also on the temperature (ductile-brittle transition) of the material. The effect of temperature on the nature of the fracture is of considerable importance. Many steels exhibit ductile fracture at elevated temperatures and brittle fracture at low temperatures. The temperature above which a material is ductile and below which it is brittle is known as the ductile–brittle transition temperature (DBTT), nil ductility temperature (NDT), or nil ductility transition temperature. This temperature is not precise, but varies according to prior mechanical and heat treatment and the nature and amounts of impurity elements. It can be determined by some form of drop-weight test (for example, the Charpy or Izod tests).
To minimize neutron fluence:
- Radial neutron reflectors are installed around the reactor core. Neutron reflectors reduce neutron leakage and therefore they reduce the neutron fluence on a reactor pressure vessel.
- Core designers design the low leakage loading patterns, in which fresh fuel assemblies are not situated in the peripheral positions of the reactor core.
If the metal is heated to elevated temperatures after irradiation (a form of annealing), it is found that the strength and ductility return to the same values as before irradiation. This means that radiation damage can be annealed out of a metal.
See also: Ductile-brittle Transition Temperature
See also: Irradiation Embrittlement
See also: Thermal Annealing
Reactor Vessel Material Surveillance Program
Reactor vessel surveillance programs provide information on the effect of radiation on vessel materials under operating conditions. Reactor vessel surveillance program utilizes capsules located on the vessel wall directly opposite the center of the core. The capsules contain reactor vessel steel specimens obtained during vessel fabrication and are withdrawn periodically from the reactor vessel. The surveillance capsules must be located near the inside vessel wall in the beltline region so that the material specimens duplicate, to the greatest degree possible, the neutron spectrum, temperature history, and maximum neutron fluence as experienced at the reactor vessel’s inner surface. A specimen capsule containing specimens for use in Charpy V-notch, tensile, and fracture mechanics tests can be removed from the reactor during normal refueling periods.
The Charpy V-notch (CVN) technique is most commonly used. The Charpy V-notch test uses a notched sample of defined cross-section. For these dynamic loading conditions and when a notch is present, we are using notch toughness. Charpy and Izod impact tests are used to measure this parameter, which is important in assessing the ductile-to-brittle transition behavior of a material. Similarly as for tensile toughness, notch toughness is measured in units of joule per cubic metre (J·m−3) in the SI system, but in this case we are measuring the area at the notch position.
There can also be special dosimeters, including pure nickel, copper, iron, aluminum-cobalt or uranium-238, which can be placed in spacers specially drilled to contain the dosimeters.
According to 10 CFR 50 Appendix H, no material surveillance program is required for reactor vessels for which it can be conservatively demonstrated by analytical methods applied to experimental data and tests performed on comparable vessels, making appropriate allowances for all uncertainties in the measurements, that the peak neutron fluence at the end of the design life of the vessel will not exceed 1017 n/cm2 (E>1 MeV).
Special Reference: NUREG-1511, Reactor Pressure Vessel Status Report. U.S. Nuclear Regulatory Commission, Washington, DC, 1994.