LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor

The MIT Research Reactor (MITR) is in the process of converting from the current 93%-enriched U-235 highly-enriched uranium (HEU) fuel to the low enriched uranium (LEU, <20%-enriched U-235) fuel, as part of the global non-proliferation initiatives. A high-density, monolithic uraniummolybdenum (U-10Mo) fuel matrix is chosen. The fuel element design is changed from 15-plate finned HEU fuel to 19-plate unfinned LEU fuel with the same geometry. The reactor power increases from 6.0 MW to 7.0 MW thermal, and primary coolant flow rate increases from 2000 gpm to 2400 gpm. Detailed analyses were completed for initial LEU core with 22 fuel elements, and demonstrated both neutronic and thermal hydraulic safety requirements are met throughout equilibrium cycles. An alternative conversion strategy is proposed which involves a gradual transition from an all-HEU core to an all-LEU core by replacing 3 HEU fuel elements with fresh LEU fuel elements during each fuel cycle. The objectives of this study are to demonstrate that the primary coolant system can be safely modified for 2400 gpm operation, and to perform steady-state and loss-of-flow (LOF) transient thermal-hydraulic analyses for the MITR HEU-LEU transitional mixed cores to evaluate this alternative conversion strategy. The primary technical challenge for the 20% increase in primary flow rate with existing piping system is flow-induced vibration. Several experiments were performed to measure and quantify vibration acceleration and velocity on three main hydraulic components to determine if higher flowrates cause excessive vibration. The test results show that the maximum vibration velocity is 9.70 mm/s, the maximum vibration acceleration is 0.98 G at the current flow rate 2000 gpm and no significant spectral change in the vibration profile at 2550 gpm. Therefore, it can be concluded that the existing piping system can safely support 2400 gpm primary flow operation. Thermal hydraulics analysis was performed using RELAP5 MOD3.3 code and STAT7 code. The MITR transitional mixed core input models were constructed to simulate the reactor primary system. Two scenarios, steady-state and loss-of-flow transient were simulated at power level of 6 MW. RELAP5 results show that during steady state, there is significant safety margin (> 10 °C) to onset of nucleate boiling for both HEU and LEU fuel. The maximum core temperature occurs at HEU fuel in Mix-core 3, the maximum wall temperature reached was 89 °C. During the LOF transient case, the result shows that The HEU fuel element is more limiting than the LEU in transitional cores. Nucleate boiling is predicted to occur only in the HEU hot channel during the first 50 seconds after the pump coastdown. The peak cladding temperatures are much lower than the fuel temperature safety limit of UAlx fuel plates, which is 450 °C. From the STAT7 calculation results, the operational limiting power at which onset of nucleate boiling (ONB) occurs in all cases show significant margins from the Limiting System Safety Setting (LSSS) over-power level. The lowest margin for LEU element during the mixed core transition is at Mix-7, 11.43 MW with a 4.03 MW power margin. For the HEU element, the lowest margin during the transition is at Mix2, 8.51 MW with a 1.11 MW power margin. The location at which ONB is always expected to occur is F-Plate Stripe 1 and 4 for the LEU fuel element; side plate for the HEU fuel element with the HEU element is always more limiting.

Name thesis:degree_name

Master

Department dc:contributor.department

Massachusetts Institute of Technology. Department of Nuclear Science and Engineering

Grantor dc:publisher

Massachusetts Institute of Technology

Year dc:date.issued

2022

Author dc:creator

• Zhao, Yinjie

Advisor dc:contributor.advisor

• Hu, Lin-Wen

Statement dc:rights

• In Copyright - Educational Use Permitted
• Copyright MIT

Licence dc:rights.uri

• http://rightsstatements.org/page/InC-EDU/1.0/

Handle dc:identifier.uri

https://hdl.handle.net/1721.1/147212

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/147212