Elliott, Alicia M. | Search NASA

DOE OSTI2019

CRUD Source Term Assessment and Development for VERA

In pressurized water reactors, corrosion of primary loop materials is an ongoing process, as primary loop components (steam generator, hot leg, cold leg, and core) are constantly exposed to the coolant. As a result of this corrosion, primary loop materials are released into the coolant in the form of aqueous metal ions and particulate oxides. These corrosion products are the source of materials that form crud deposits on fuel rod clad surfaces. The mass of corrosion products available to create crud will limit the rate and extent of crud buildup that occurs. This is commonly referred to as the crud “source term”. The major contribution to the source term mass comes from the steam generator, due to the large surface area exposed to the coolant. Steam generators are not identical; the wetted area exposed to the coolant and the alloy used in the steam generator tubes vary. Different alloys corrode at different rates; Inconel 690 corrodes at approximately one-third of the rate of corrosion of Inconel 600. The mass of corrosion products released from the steam generator also varies by the area exposed to the coolant. The release of these corrosion products depends on the local environmental factors within the primary loop, including temperature, pH, local fluid conditions, electrochemical potential, and thermodynamic solubilities of the species present within this system. As such, each PWR will have a different crud source term based on these plant-specific parameters. In order to accurately model and predict crud growth and issues that arise from the presence of crud (CIPS, CILC), the source of these corrosion product materials must be accurately modeled in a system mass balance. These models should be mechanistic, representative of the underlying physics phenomena, and account for the effects of the plant-specific design and operating parameters, including the alloys used, local temperatures, and fluid flow conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Thermal Desorption of Tritium from Fluoride Salt-Cooled Reactor Materials

The fluoride-salt-cooled high-temperature reactor (FHR) is a high-temperature fission reactor concept involving a liquid fluoride salt coolant, coated particle fuel, and enhanced passive safety mechanisms. Utilizing a low-pressure coolant and high temperature fuel allows for smaller reactor buildings and use of gas Brayton power cycles for higher efficiency in electricity production. One of the fluoride salts being considered for the FHR coolant is 2LiF - BeF{sub 2} (flibe). Tritium is generated frequently in flibe from thermal neutron transmutation of Li-6 and Be-9. Tritium retention in irradiated structural materials presents a challenge to the design of the FHR; high neutron fluence can induce significant material damage and alteration of microstructures to trap tritium in the structure, particularly in carbon-based materials (such as steels). Investigation of tritium transport in flibe salt and structural materials is also critical to evaluating mass balance of tritium in a system for quantifying leakage of tritium to the environment. It is important to understand the uptake of tritium in these proposed core materials since the transport of tritium within an FHR is of critical concern for operational safety due to corrosion, and to prevent tritium release into the environment. The presence of TF corrodes stainless steel by oxidizing chromium. Tritium gas (T{sub 2}) easily diffuses through reactor materials and can be released into the environment through the primary piping and heat exchangers. There is great interest in identifying materials that are compatible with the FHR environment that can be used to getter tritium from the salt, or to inhibit the diffusion of tritium to prevent its release into the environment. As part of the FHR Integrated Research Project (IRP), capsules containing proposed core materials immersed in flibe salt were irradiated at 700 deg. C for 1000 hours (FS-1) and 300 hours (FS-2) in the MIT Research Reactor (MITR). During irradiation, tritium is produced in the salt, and then interacts with the other materials in the capsule. Measurements of the effluent gas during the FHR IRP irradiation experiments indicated that less than 1% of the tritium generated in the salt escaped in the gas phase. It is assumed that most of the tritium was captured, either via diffusion or surface adsorption, on the materials inside the capsule. Tritium frequently adsorbs to sample surfaces in the form of HTO, which requires relatively low temperatures to release (on the order of 100 deg. C). In certain cases, tritium can be bound in the bulk of materials from diffusion or neutron transmutation of lithium and beryllium in the structural materials. Accurate measurement of very small concentrations of hydrogen is difficult; by taking advantage of the radioactive decay of tritium (which emits low-energy beta radiation), several methods become available to detect and quantitatively measure the tritium. The most sensitive method is liquid scintillation counting (LSC), which requires the tritium to be suspended in optically-transparent liquid media that is compatible with scintillating solution, such as water. This measurement technique requires the tritium to be extracted from the material, which can be achieved through leaching into water, thermal desorption, or complete destruction of the sample by acid digestion or pyrolysis. A less destructive method is preferred in order to preserve the specimens, reduce the generation of radioactive mixed wastes, and minimize the carryover of other activation products. For total thermal desorption of tritium from the materials, the tritiated materials must be heated to temperatures in excess of 300 deg. C. Heating the specimens above the irradiation temperature will re-mobilize the tritium and allow it to be captured from the gas stream directly into LSC media. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗