DOE-NE is supporting the development of FHRs primarily through its nuclear energy university program as a small-scale university endeavor.
The central development challenges for FHRs are twofold: (1) maturing the technology sufficiently to enable deployment and (2) achieving a sufficiently low cost of power generation to become a preferred energy provider. The leading FHR development topics that remain to be addressed are discussed briefly here. The Fluoride-Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap provides additional information on the development path.
Tritium control
Tritium control is an important issue for FHRs because tritium is the only radionuclide that has potential for significant release under normal operating conditions and without failed fuel. The large contact surface area and thin walls of the heat exchanger tubes means that heat exchangers will be the primary release pathway. Demonstrating the recently invented tritium stripping technology is key to resolving this issue.
Lithium-7 cost
Individual units of large-scale FHR nuclear power plants are anticipated to use a few hundred tons of isotopically selected (~99.995%) 7Li. Therefore, for FHRs to be economically preferable, a reliable, cost-effective supply of 7Li is required. Lithium isotopes can be separated by conventional chemical technologies with well-understood production volume–cost scaling relationships. Although several alternative techniques have potential for separating lithium isotopes on an industrial scale and at reasonable cost, it is not yet possible to specify a technically preferred lithium isotope separation technique.
Licensing
Licensability is a key element of any reactor development effort. While the inherent safety features of FHRs offer a compelling case for safe operation to support a licensing basis, an FHR safety demonstration and licensing effort will still need to include a number of elements:
· Define and support the safety basis
o Define FHR-specific GDCs
o Establish a quality assurance program
o Establish response to transients and accidents
§ Identify accident-initiating events
§ Select licensing basis events
§ Categorize licensing basis events
§ Identify and classify the SSCs
o Determine the adequacy of the reactor design and supporting information needs
§ Perform a regulatory gap analysis
§ Implement an FHR-specific fuel qualification program
§ Develop severe accident evaluation capability
o Determine the adequacy of codes and standards
o Define a program to address the hazards associated with the presence of beryllium in FHR coolant
· Demonstrate the adequacy of models
· Demonstrate the adequacy of design
Fuel development and qualification
FHRs will use coated-particle fuel. DOE-NE is testing TRISO coated-particle fuel as part of its HTGR development efforts. Identical TRISO particles are directly applicable to FHRs. The initial TRISO fuel loads for first-generation FHRs will cost substantially more than LWR fuel pellets. TRISO fuel is not currently manufactured at the commercial scale. Consequently, the cost savings resulting from manufacturing scale-up and automation cannot be reliably estimated at present. FHRs, like LWRs, are thermal spectrum reactors intended to run on a once-through low-enrichment uranium fuel cycle. However, FHRs will require somewhat higher 235U enrichment than that currently employed at LWRs, and modifying the existing fuel infrastructure will be expensive. Because LWRs can achieve higher burnup by using higher enrichment fuel, planning to upgrade US commercial fuel enrichment capabilities has already begun. The nonmanufacturing fuel costs for FHR TRISO are expected to be similar to those for the more highly enriched LWR pellets.
Structural alloy development and qualification
A substantial body of knowledge exists for Alloy N, the leading candidate material for FHR test reactors. Alloy N, however, is not currently approved as a material for high temperature nuclear power plants. Thus, a materials qualification effort would be necessary if Alloy N were used in any part of the containment boundary or in any other safety function. Developing a safety case for limited-term Alloy N use at temperatures less than 704°C may be possible, based upon existing data from the earlier ORNL MSR program and/or limited-term supplemental qualification testing. The maximum allowable stress for Alloy N decreases rapidly above 600°C, becoming too low for practical use above 700°C. Compared with the candidate materials typically considered for high temperature nuclear reactor construction, Alloy N has significantly lower high temperature strength. However, developing and qualifying improved performance alloys for longer-term, high stress applications will require substantial investment over a number of years. ORNL has recently begun the development of successor alloys to Alloy N. Samples of the new alloys show promise for application in high stress applications at higher temperatures along with good fluoride salt corrosion resistance. As a first step, it is recommended that early phase, long-term advanced salt-compatible alloy property measurements be performed. With this understanding, improved alloys can be developed and made available for commercial FHR deployments while a limited-term safety case for use of Alloy N at test reactors can be developed in parallel.
Continuous fiber composites
FHRs will make extensive use of continuous fiber composites (CFCs) for reactor vessel internal components. Some of these components will be large and have complex geometry (e.g., the lower core support plate). CFCs are being evaluated for in-vessel structural applications in other reactor classes (e.g., using SiC-SiC CFCs as channel boxes at boiling water reactors to minimize the core zirconium content); however, a significantly larger role for CFCs is envisioned at FHRs due to their ability to maintain their structural characteristics at high temperatures.
Contacts
David Holcomb, HolcombDE@ornl.gov, Salt Reactor Technical Lead
Jess Gehin, GehinJC@ornl.gov, ORNL Reactor Technology Lead
Gary Mays, MaysGT@ornl.gov, Advanced Reactor Systems and Safety Lead