The Reactor Physics Analysis Group functions in two major domains: 1) technical support of MITR operation and experiment and 2) cutting edge investigation of advanced reactor concepts and innovative test reactor design.

The code system used in the group and on-going research projects are presented below. For inquiries about specific project details and future collaborations, please fill out our contact form.

MITR Fuel Management and Code Development

Considering the current quarterly refueling scheme, the fuel consumption is about nine fuel elements per year. Recently, the fuel management code has been upgraded from finite-difference diffusion code CITATION (with in-house depletion solver) to state-of-the-art general purpose Monte Carlo code MCNP (with in-house depletion code package MCODE). The newer method allows more comprehensive understanding of the fission density distribution in the MITR core. It is also capable of taking into account all the flipping and rotating moves for the symmetrical MITR fuel element design. Fissile utilization can be effectively improved thanks to the on-going code development and validation.

MITR LEU Conversion

In the framework of non-proliferation policy, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors, both domestic and international, have started a program of conversion to the use of low enriched uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR. The Preliminary Safety Analysis Report (PSAR) for Nuclear Regulatory Commission (NRC) is currently under preparation.

Engineering Design and Safety Evaluations of MITR In-core Experiments

Four types of in-core experimental facilities at the MITR have been developed and demonstrated: (1) a flexible, general purpose in-core sample assembly (ICSA) for capsule irradiations at temperatures up to 900°C, (2) a very high-temperature irradiation facility (HTIF) for irradiations in the range of 1200-1400°C, (3) LWR loops that can be operated at pressurized or boiling water reactor conditions, and (4) an experimental fuel irradiation facility. Each new in-core experiment requires a safety evaluation report, indicating no reactivity, radiation, and temperature limit will be exceeded.

Fluoride-salt-cooled High-temperature Reactor (FHR)

The Fluoride-salt-cooled High temperature Reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature TRISO particle fuel. The refractory TRISO particle coating system as well as the dispersion in graphite matrix enhances safeguards and security. Comparing to the conventional high temperature reactor cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a Nuclear Air-Brayton Combined Cycle (NACC) that provides base-load and peak power capabilities. The FHR is therefore considered as an ideal candidate for the transportable reactor concept. In this context, a 20 MWth compact FHR, feasible to be transported by a truck and aiming at an 18-month once through fuel cycle, is currently under design at NRL. Demonstration attempts are being investigated to replace the MITR Fission Converter with a subcritical FHR system.

Transient Reactor

Efforts are presently underway to restart the Transient Reactor Test (TREAT) Facility, located at the Idaho National Laboratory (INL) by 2018. The TREAT Facility has historically been utilized to provide empirical data to support the comprehensive characterization of light water reactor fuel and sodium-cooled fast reactor fuel under a variety of conditions. In the framework of this restart program, NRL is leading an integrated instrumentation plan for the TREAT Facility and participating an extensive evaluation of existing TREAT Facility neutronics data according to established guidelines per the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook).