Materials and Chemistry for Molten Salt Systems: Loops and Irradiation Effects
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee
Program Organizers: Stephen Raiman, University Of Michigan; Kumar Sridharan, University of Wisconsin-Madison; Nathaniel Hoyt, Argonne National Laboratory; Jinsuo Zhang, Virginia Polytechnic Institute and State University; Michael Short, Massachusetts Institute of Technology; Raluca Scarlat, University of California, Berkeley
Tuesday 8:30 AM
March 16, 2021
Room: RM 49
Location: TMS2021 Virtual
Session Chair: Stephen Raiman, Texas A&M University
8:30 AM
Corrosion and Mass Transfer of 316H Stainless Steel in Flowing FliNaK Salt: Stephen Raiman1; Matthew Kurley2; Dino Sulejmanovic2; Scott Nelson2; James Keiser2; Bruce Pint2; 1Texas A&M University; 2Oak Ridge National Laboratory
Researchers have been using thermal convection loops (TCLs) since the 1950s. By imposing a temperature difference between a hot leg and a cold leg, TCLs can be used to recreate the mass transfer that occurs as the solubility of species in the flowing salt shifts with temperature. For this work, 316H stainless steel coupons were exposed for 1000h in a 316H thermal convection loop filled with purified FLiNaK salt. The loop was operated with a maximum hot leg temperature of 650°C, and a cold leg minimum temperature of 540°C. This talk will present the results of this most recent TCL experiment, along with TRANSFORM (TRANsient Simulation Framework Of Reconfigurable Models) modeling to better understand the properties that affect mass transfer. Historical context will also be given and future plans for material compatibility testing in flowing molten salts will be discussed.
8:50 AM Cancelled
Dutch Molten Salt Irradiation Program: Ralph Hania1; Uazir Bezerra de Oliveira1; 1NRG
NRG is the Dutch national nuclear laboratory and operates the High Flux Reactor in Petten. Within the Dutch national R&D program innovative nuclear systems are studied. MSR irradiation projects are ongoing under the names SALIENT, SAGA and ENICKMA. The SALIENT irradiations are in-core capsule irradiations of fluoride fuel salt samples, performed to gain experience with the handling, irradiation, examination and waste treatment of molten salts. Goal is to investigate fission product migration, noble metal plate-out and salt-capsule interactions. SAGA is a gamma irradiation of fresh and solid fluoride salts at near room temperature. At these low temperatures, fluoride salts are susceptible to radiolysis, resulting in the production of fluorine gas, to be quantified by SAGA. Finally, the ENICKMA irradiations study the degradation of nickelbase alloys under irradiation. In this contribution we will give an overview of the Dutch molten salt program with focus on the most recent results.
9:10 AM
Design of Molten Salt Static Corrosion Experiments to Predict Phenomena Relevant to Corrosion in Non-isothermal Nuclear Reactor Salt Loops: Raluca Scarlat1; 1University of California, Berkeley
Corrosion of metal alloys in the salt loops of a molten salt (MSR) or a salt-cooled nuclear reactors (FHR) is continuously driven by the presence of a thermal gradient coupled with mass convection. This system-level driving force creates a local thermodynamic driver for corrosion of metal piping and components. The question of whether similar corrosion conditions can be generated in static corrosion capsules will be discussed. Fission and neutron activation reactions, and inadvertent ingress of materials into the coolant can have an oxidizing effect on the salt and are another driver of corrosion, which is generally mitigated by redox control. How these effects could or should be replicated in static corrosion capsules will be discussed. Examples of redox control and redox measurements will be demonstrated, and a discussion of the limitation of static corrosion capsules will be provided.
9:30 AM
Structural Health Impacts Due to Exposure of Irradiated Molten Chloride Salts: Nora Dianne Ezell1; Stephen Raiman2; Joel McDuffee1; Matt1; 1Oak Ridge National Laboratory; 2ORNL
Understanding the effects molten salts have on the infrastructures they encounter is critical for structural health monitoring, planned maintenance, and safe operation. While there are many corrosion studies at elevated temperatures, the temperature is not the only hazard in a molten salt reactor. ORNL executed a small-scale proof-of-concept irradiation to understand the effects on corrosion development due to radiation. The goals of this irradiation were to demonstrate the possibility of a safe and effective irradiation experiment and to obtain preliminary corrosion data.This work discusses the PIE results from a chloride salt irradiation experiment executed at the Ohio State Research Reactor in 2018. While this is a proof-of-concept irradiation, many lessons were learned that will impact the design of the next, larger-scale follow-on irradiation. This document will discuss the future plans and suggestions for increasingly more complex irradiation experiments.
9:50 AM
Alloy Compatibility in Flowing Cl and F Salts: Bruce Pint1; Dino Sulejmanovic1; J. Kurley1; Stephen Raiman1; 1Oak Ridge National Laboratory
There is considerable interest in molten salts for concentrating solar power (CSP) and nuclear applications. The most important assessment of salt compatibility is in a flowing, non-isothermal experiment and thermal convection loops (TCLs) have been used for decades to provide compatibility data. For the CSP application with commercial MgCl2-KCl-NaCl salt, small mass changes were found with both purified (>500 ppm O) salt and when the salt was only dried. Classic dissolution-precipitation behavior was observed after 1000 h operation for both NiCr alloys 600 and C276 in an alloy 600 TCL with a peak temperature of 700°C and no degradation of the room temperature tensile properties was observed. For a nuclear application, a TCL is being constructed from type 316H stainless steel for operation with FLiBe salt and a peak temperature of 650°C is targeted. Results will be presented after 1000 h operation.
10:10 AM
Chemical Effects of Ionizing Radiation on Molten Salt Systems: Simon Pimblott1; Ruchi Gakhar1; Gregory Horne1; Kazihiro Iwamatsu2; Alejandro Ramos3; Jay LaVerne3; James Wishart2; 1Idaho National Laboratory; 2Brookhaven National Laboratory; 3University of Notre Dame
Ionizing radiation in the molten salt reactor environment will drive chemical changes in the composition of the salt and induce additional pathways for corrosive degradation of reactor components. Although the products of radiolysis are largely expected to recombine in the high-temperature molten salts, even a small mechanistic leakage out of the recombination cycle will result in substantial formation of permanent products, such as halogen gas (F2 or Cl2 depending on the salt) or metal particles that may deposit on surfaces in the reactor circuit. Thus, it is important to understand the initial products of molten salt radiolysis and their reaction kinetics with solutes. MSR fuel and corroded infrastructure materials are full of potential scavengers for these initial products. State-of-the-art steady-state and pulse radiolysis techniques are being deployed to understand the radiation-induced chemistry of molten chloride salt systems and to develop a predictive multi-scale radiation chemical kinetics model.
10:40 AM
Microstructural Characterization of Grain Boundaries in Hastelloy N Corroded in Molten FLiBe Salt under Neutron Irradiation: Guiqiu Zheng1; David Carpenter1; 1Massachusetts Institute of Technology
The grain boundary (GB) of structural alloys plays an important role in the evaluation of corrosion resistance, radiation tolerance, thermal stability, and mechanical strength during their applications in extreme environments such as molten salt reactors. The degradation of GB strength induced by molten salt corrosion and neutron irradiation is one of the critical factors causing intergranular cracking of structural alloys, which dramatically shortens the service life of reactors. Hastelloy N and 2LiF-BeF2 (FLiBe) are being considered as the primary structural alloy and coolant for fluoride salt-cooled high-temperature reactors (FHRs). To evaluate the performance of Hastelloy N in simulated FHRs environment, Hastelloy N coupons were irradiated in molten FLiBe salt in both nickel-lined and graphite crucibles in the MIT Research Reactor at 700°C for 1000 hours. In this study the GB microstructure and microchemistry of irradiated Hastelloy N were characterized using the state-of-the-art instruments at the Idaho National Laboratory.
11:00 AM
Exploration of the Corrosion Morphologies of Ni-Cr Alloys in Molten Fluoride Salts with/without Radiation: Weiyue Zhou1; Yang Yang2; Miaomiao Jin3; Andrew Minor2; Michael Short1; 1Massachusetts Institute of Technology; 2Lawrence Berkeley National Laboratory; 3Idaho National Laboratory
Corrosion of structural materials in molten fluoride salts proceeds mainly via selective leaching of Cr from the alloys into the salts. The resulting corrosion morphologies can be entirely different despite identical Cr removal quantities, therefore affecting the quantification of the severity of corrosion attack. The morphologies of the corrosion induced microstructures relate directly to the processes of Cr transport and redistribution of other atoms accordingly. In this sense, parameters such as temperature, Cr concentration, salt corrosivity, radiation, etc., will all directly and/or indirectly influence morphology. Here we present a mechanistic mapping of molten salt corrosion versus Cr concentration and temperature in binary Ni-Cr alloys. The influence of other parameters like salt corrosivity and radiation on the morphologies will also be commented upon.
11:20 AM
Release Behavior of Tritium Generated inside FLiNaBe by Thermal Neutron: Kazunari Katayama1; 1Kyushu University
Florine molten salts such as FLiBe and FLiNaBe are promising liquid blanket materials for fusion reactors. The understanding of tritium behavior generated by the nuclear reaction of Li with neutron is important from a viewpoint of safety. The information of tritium behavior in the florine molten salts is also useful for molten salt fission reactors. In this study, we observed the release behavior of tritium generated inside FLiNaBe by neutron irradiation at Kyoto University Research Reactor. The solid-state samples of FLiNaBe was irradiated by thermal neutrons with the total fluence of 1.7×10^15cm-2. After neutron-irradiation, the samples were put into a Mo crucible installed in a stainless steel tube and heated to 700°C in Ar gas flow. The release ratio of each chemical form was approximately TF:HT:HTO=15:65:20. Initially, the most of tritium was released as HT but eventually, the release rate of TF was increased after long time heating.