Liquid-fueled molten salt reactors (MSRs) present unique challenges for traditional methods of nuclear material accounting and control (NMAC), due in large part to the novel fuel form. With at least 19 distinct liquid-fueled MSR designs in various stages of development worldwide, developers planning to submit license applications to their national regulatory body and the International Atomic Energy Agency (IAEA) need approaches and technologies to quantify nuclear material within future MSR facilities. In response to this need, an effort at Oak Ridge National Laboratory is underway to develop, design, and test hardware components of a laboratory-scale prototype system to monitor initial and makeup fuel salt additions in a liquid-fueled MSR. A methodology for surveying the various measurement techniques that could be utilized in the system has been developed and implemented. The methodology includes the identification of nine figures of merit (FOMs) designed to be evaluated in the context of establishing safeguards for a liquid-fueled MSR. These FOMs include cost, measurement uncertainty, measurement time, operator burden, and technology readiness level. The evaluation leveraged prior work, literature review, and expert feedback for 18 techniques that could provide relevant information toward the goal of quantifying the total uranium and 235U fraction in fresh fuel salt as it is added to an MSR core. A final system was selected based on this survey and specific use scenarios. The outcome of this work will be a prototype design for an MSR feed monitoring system. The developed system will provide in situ, unattended quantification of fresh fuel additions to reactor containment, which will meet the ultimate goals of (i) providing the capability to detect diversion of declared fresh fuel, (ii) ensuring that the facility is not being misused (e.g., by introducing undeclared fertile material into the reactor for purposes of fissile material production), and (iii) when combined with operational data, bounding the quantities of nuclear material that may be present in irradiated salt (to help reduce the impact of uncertainties on measurements of irradiated salts). The goal of this project is to provide a system that can support MSR developers as an element of their MC&Aprogram or the IAEA for independent verification of nuclear material quantities added to an MSR. The information gained from the system will provide a scientific basis and proposed technical system for making measurements of fresh feed in liquid-fueled MSRs for safeguards and security purposes.
Year
2024
Abstract