During the service life of a reactor pressure vessel (RPV) in a Light Water Reactor (LWR), nascent hydrogen produced at the steel-water interface at the inner wall is absorbed through the austenitic stainless steel cladding into the RPV steel, and diffuses through the metal until discharged to atmosphere on the outer surface. The aim of this report is to assess whether this hydrogen will increase the hardness, reduce the ductility, or facilitate crack growth in the RPV beyond the changes induced by radiation alone.
The approach used has been to perform a literature review and utilise the information reviewed within different damage scenarios. The review provides a framework for the literature on irradiated material by considering first the routes whereby hydrogen may be introduced into a component or sample, and summarising the literature on hydrogen effects on different mechanical properties in iron and low alloy steels in the absence of radiation. This illustrates the amount of hydrogen required to affect the outcome of different types of mechanical test (tensile, Charpy, fatigue, fracture toughness). It also shows that mobile hydrogen is more significant than trapped hydrogen in affecting mechanical properties.
Reviewing the limited number of observations of neutron irradiated and hydrogenated steels shows that hydrogen embrittlement and radiation damage are generally additive. Most commonly, the ductility or toughness of irradiated low alloy steels follows a tanh-type dependence on hydrogen, with little effect over some hydrogen range, followed by a drop to a low level at high hydrogen content. The hydrogen level below which minimal effects are observed may depend on dose and temperature, but is of the order 1-2 wt. ppm. Neutron irradiation increases the amount of hydrogen found in low alloy steel samples by increasing the number of trap sites. This complicates the interpretation of mechanical property changes in hydrogenated, irradiated, low alloy steel. Assessments of the mobile hydrogen content require input from modelling. Calculations show that hydrogen dissolved in the coolant is the main source of hydrogen in the vessel wall. Historically, a level of 2 wt. ppm hydrogen has been considered a reasonable, conservative estimate for the RPV wall. For a modern 200 mm thick low alloy steel vessel with 7 mm thick austenitic clad operating at 300°C, a conservative estimate of the hydrogen is 0.45 wt. ppm at the clad-base metal interface, falling to zero at the outer vessel wall. A best estimate is 0.02 wt. ppm at the clad-base metal interface. Even the conservative estimate would induce only small effects on the average tensile and fracture toughness properties at room temperature, and none at the operating temperature. Despite unavoidable uncertainties in some of the input parameters, calculations of the hydrogen buildup in flaws within the RPV indicate that the pressures will be too low to induce flaw extension.
|Title of host publication
|Potential Effects of Hydrogen-Related Degradation on Reactor Pressure Vessel Materials (MRP-431): Review of Current State of Understanding
|EPRI - Electric Power Research Institute
|Number of pages
|Published - 1 Dec 2018
|MRP-431 - 3002013225