Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

Abstract

This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. 9. References Appendix 1. Pressure drop correlation Appendix 2. Impact of the number of fuel pins per assembly on thermal hydraulics Appendix 3. Examples of hot channel thermal calculations Appendix 4. Review of other fast test reactors Appendix 5. Approach to determine the reactivity coefficients Appendix 6. Detailed results from parametric study Fuel (60 to 100 cm) Fission gas plenum (same height as fuel) Reflector (50 cm) Reflector (50 cm) Driver Fuel Control Rod Safety Rod Experiment Reflector Shield 4 2.2 Design Constraints for Normal Operation In order to ensure the reactor is operated safely and efficiently, the design requirements discussed in this section were considered for this trade study. These values are solely used in the scope of the trade study and may be reviewed and adjusted later.  Fuel smeared density: 75% theoretical density  The core reactivity is controlled with six primary control rods and three secondary safety rods. The rod locations are shown in Figure 2.2; these positions were arbitrarily selected to carry out the trade study and will need to be adjusted based on control and safety considerations (see Section 5.1).  Sodium inlet and outlet temperatures: 350C and 500C, i.e., representative values for sodium-cooled fast reactors. These are temperatures assumed only for this trade study. More detailed thermal and safety analyses will be required to converge on optimized values. Increasing this inlet-outlet sodium T could be beneficial to increase the fast flux level by allowing increasing power density without increasing sodium velocity.  Peak cladding temperature  650C. The peak cladding temperature is mostly dependent on the sodium outlet temperature. At this stage, no detailed thermal calculations and no orificing optimization have been performed, but based on experience, it is known that a 500C average sodium outlet temperature will ensure the peak cladding temperature requirement is met, provided the fuel assembly is properly designed.  Peak fuel temperature  1,121C for U-20Pu-10Zr and  1,248C for U-10Zr. The peak linear power assumed for this parametric study (450 W/cm, see below) should be low enough to meet this requirement with some...

"Advanced" isn't Always Better: Assessing the Safety, Security, and Environmental Impacts of Non-Light-Water Nuclear Reactors

“…15 This burnup may actually be too high. The DOE has limited the average discharge burnup for PRISM-type metal fuel for the very similar Versatile Test Reactor to about 6 percent (approximately 60,000 MWd/MTHM) in order to remain within the envelope of the current fuel qualification database (Youinou et al 2020). 16 This is different from the Very High Temperature Reactor, or VHTR, a less mature design originally included in the Generation IV program that would have an outlet temperature up to 950°C.…”