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ORNL and Shanghai Institute of Applied Physics in CRADA for development of fluoride salt-cooled high-temp reactors

Oak Ridge National Laboratory and the Shanghai Institute of Applied Physics (SINAP) are engaged in a Cooperative Research and Development Agreement (CRADA) focused on accelerating scientific understanding and technical development of salt-cooled reactors, specifically fluoride salt-cooled high-temperature reactors (FHRs). The project will draw on ORNL’s expertise in fuels, materials, instrumentation and controls, design concepts, and modeling and simulation for advanced reactors, as well as the lab’s experience in the design, construction and operation of the Molten Salt Reactor Experiment, the only molten salt reactor ever built. (Design began in 1960, construction started early in 1962. The 7.4 MWth test reactor operated successfully from 1965 to 1969.)

Representatives from the Oak Ridge National Laboratory (ORNL) and the Shanghai Institute of Applied Physics (SINAP) are meeting at ORNL this week; SINAP staff members will describe their plans for building the first salt-cooled test reactor, and the two sides will begin planning the next steps in the shared research project.

FHRs. FHRs feature low-pressure liquid fluoride salt cooling; solid coated particle fuel; carbon-based neutron moderation; fully passive decay heat rejection; and a high temperature power cycle. FHRs have the potential to economically and reliably produce large quantities of electricity and high temperature process heat while maintaining full passive safety, according to ORNL. Leveraging the inherent safety characteristics of FHRs avoids the need for expensive, redundant safety structures, systems, and components (SSCs), providing the opportunity for substantial cost reduction.

The safety functions in a nuclear reactor must:

  • control the reactivity
  • cool the fuel
  • prevent the release of radiation

Part of the value proposition for FHRs is that they potentially can perform these functions at substantially lower cost than current light water reactors (LWRs) through reliance on the inherent characteristics of the fuel and coolant.

As an example, one preferred primary fluoride salt coolant—27LiF-BeF2 (FLiBe)—offers desirable neutronic properties, good hydraulic performance, and very low activation. The boiling point of FLiBe is more than 1400°C, and its volumetric heat capacity at 700°C (4.67 J/cm3-K) is comparable to that of water at 100°C (4.04 J/cm3-K).

FHRs exhibit no cliff-like phenomena (such as departure from nucleate boiling or substantial coolant pressure rise with temperature) in their temperature or heat transfer characteristics. Consequently, their requirements for reactivity control are much less stringent than those of LWRs.

Additionally, as a high temperature reactor class, FHRs can efficiently generate electricity and provide the energy for high temperature industrial processes—including the production of hydrocarbon fuel. A 700°C output temperature enables a thermal cycle efficiency of 45% as compared with the 33% efficiency typical for LWRs. Moreover, unlike gas-cooled reactors, nearly all of the energy is available at the high temperature, improving the capability of FHRs to support high temperature thermal processes.

Moreover, high temperature operation increases FHR compatibility with dry cooling. No FHR has been built, and significant technical hurdles must be overcome before FHRs can realize their potential. However, Oak Ridge notes, no concept viability issues beyond the grasp of an appropriate research, development, and demonstration program have been identified.

Issues that need to be addressed include:

  • Tritium control. Tritium is the only radionuclide that has potential for significant release under normal FHR 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. A recently invented tritium stripping technology is key to resolving this issue, according to ORNL scientists.

  • 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; a reliable, cost-effective supply of 7Li is thus necessary. 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.

  • Fuel development and qualification. FHRs will use coated-particle fuel. The US Department of Energy (DOE) Office of Nuclear Energy is testing TRISO (tristructural isotropic) coated-particle fuel as part of its HTGR development efforts. Identical TRISO particles are directly applicable to FHRs.

    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; 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, according to ORNL, 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 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, ORNL is recommending 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.

  • Licensing. Licensability is a key element of any reactor development effort.

A number of organizations have been developing FHR reactor concepts:

  • ORNL’s two reactor concepts in the FHR class are the 1500 MWe advanced high temperature reactor (AHTR) and the 125 MWt small modular advanced high temperature reactor (SmAHTR).

  • MIT is investigating a fluoride-salt-cooled high temperature test reactor (FHTR).

  • Georgia Tech is investigating a liquid salt cooled reactor (LSCR).

  • UC Berkeley is developing a pebble-bed fluoride-salt-cooled high temperature reactor called the Mark 1 (Mk1 PB-FHR).

  • SINAP is pursuing the development of its first test FHR, designated the Thorium Molten Salt Reactor – Solid Fuel 1 (TMSR-SF1).

  • In January 2015, Canada-based Terrestrial Energy (TEI) entered into an initial collaboration with ORNL to advance the Terrestrial’s Integral Molten Salt Reactor (IMSR) to the engineering blueprint stage, expected in late 2016. TEI’s IMSR is a small modular design, with models ranging from 80 MWth to 600 MWth.

The ORNL-SINAP CRADA. The CRADA evolved from US–China interactions under a Memorandum of Understanding between the US Department of Energy (DOE) and the Chinese Academy of Sciences (CAS) on “Nuclear Energy Sciences and Technologies Cooperation.”

DOE is responsible for developing nuclear energy concepts with the potential to provide significant safety and economic improvements over existing reactors, a mission carried out by the Advanced Reactor Technologies Program in DOE’s Office of Nuclear Energy.

The CAS has initiated a large FHR development program with similar objectives and has provided resources for research, technology development, design and construction of an FHR test reactor in China. This initial test reactor will have a maximum thermal power of 10 megawatts. A second, 100-megawatt test reactor is also planned. Both FHR test reactors will use low-enrichment uranium fuel.

SINAP is leading the Chinese effort on behalf of CAS to develop FHRs to supply process heat and electricity to China’s growing economy, especially in regions with limited water. The Institute is engaged in the full spectrum of activities necessary to evaluate, design, license, construct, and operate FHR test reactors, with several hundred staff devoted to its FHR development program. The United States has the most experience worldwide with technologies applicable to FHRs due to its historic and long-term investments in advanced nuclear reactors (in particular ORNL’s experience in molten salt reactor development and demonstration), advanced materials, and coated particle fuel.

The CRADA is a formal agreement between SINAP and ORNL, covering work approved by DOE as consistent with the DOE’s overall mission for advancing reactor technology. CAS is providing the entirety of CRADA funding, with an estimated $5 million a year. ORNL expects to engage other US institutions for additional support. The collaborations under the new agreement are authorized for 10 years. The CRADA focuses on resolving the technology issues associated with design, construction, and operation of FHRs, and does not include activities related to fuel reprocessing or fissile material separation.

Development of advanced reactors in the United States will benefit through access to the information and experience produced by the large and rapidly advancing Chinese program. China will benefit by leveraging access to US advanced reactor capabilities, facilities, and experience.

The United States also has substantial manufacturing skills with high-value specialized materials and components necessary to construct advanced reactors. Cooperation between ORNL and SINAP will help to ensure that US manufacturers have the opportunity to compete in future salt-cooled reactor markets.



Nick Lyons

MSRs can't get here fast enough.

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