MITR In-Core Experiments

MITR In-Core Experiments

*Experiments funded by Nuclear Science User Facilities (NSUF)
**Experiments funded by other DOE programs, such as NERI, NEET, NEUP, GAIN, SBIR, STTR etc.

Experiment Title Environment In-Service Purpose of Experiments
Fluoride Salt 4 (FS-4) Inert gas, 700 °C 2018 Tritium permeation measurements with and without the presence of liquid lithium flouride-beryllium fluoride salt, and test of tritium barrier coating for the High Tmeperature Fluoride Salt Reactor. Also demonstrated the use of a "graphite vertical" position outside the core for high temperature neutron irradiations.
TREAT Instrumentation Tests Atmospheric air, 50 °C, Reactor power <100 kW, steady state and transient power conditions 2017, 2018 Transient Test Reactor (TREAT) instrumentation tests: neutron and gamma detectors including newly developed micro-pocket fission detectors with flux wires for benchmarking
Advanced Cladding Irradiation (ACI)-Accidental Tolerant Fuel (ATF) Phase 2** Water, 290 °C 2017-2018 Corrosion, bond and creep testing of SiC/SiC composite, coated Zircaloy and 3D printed alloy materials for accident tolerant fuel/advanced clad applications in PWRs - Westinghouse-led program
ULTRA 2* Inert gas, 450 and 750 °C 2017 Performance testing of multiple ultrasonic and fiber optic sensors under irradiation in inert gas
ACI-Shared Irradiation** Water, 290 °C 2017 Corrosion and bond testing of samples duplicating the HYCO test matrix (samples with n and γ exposure, γ exposure, and coolant exposure only) plus samples from Ceramic Tubular Products
Hybrid Composite (HYCO) ATF Clad Irradiation Inert gas, 350 °C 2016 Irradiation of a large set of candidate high performance LWR clad materials for post-irradiation property testing
Fluoride Salt 3 (FS-3)** Inert gas, 700 °C 2016 Tritium transport measurements on SiC/SiC and graphite samples irradiated in liquid lithium fluoride-beryllium fluoride salt for the High Temperature Fluoride Salt Reactor. Samples from U.S. DOE program and Chinese Academy of Sciences
In-core Crack Growth Monitoring* Water, 290 °C 2016 Demonstration of actively loaded crack growth sample with real time crack length monitoring
Fluoride Salt 2 (FS-2)** Inert gas, 700 °C 2014 Corrosion and tritium transport measurements on SiC/SiC and metal coupons irradiated in liquid lithium fluoride-beryllium fluoride salt for the High Temperature Fluoride Salt Reactor IRP
ACI-Accidental Tolerant Fuel Phases 1A, 1B (ATF)** Water, 290 °C 2014-2016 Corrosion and bond testing of SiC/SiC composite, coated Zircaloy and “hybrid” clad samples for accident tolerant fuel applications in PWRS - Westinghouse-led program
ACI-BWR** Water, 290°C 2013-2014 Corrosion and creep testing on SiC/SiC BWR channel box coupons under BWR core conditions
ULTRA* Inert gas, 430 °C 2013-2014 Performance testing of multiple magnetostrictive and piezoelectric ultrasonic sensors under irradiation in inert gas
Fluoride Salt 1 (FS-1)** Inert gas, 700 °C 2013 Corrosion and tritium transport measurements on SiC/SiC and metal coupons irradiated in liquid lithium fluoride-beryllium fluoride salt for the High Temperature Fluoride Salt Reactor IRP
LUNA 2 (In-Core Sample Assembly - ICSA)** Inert gas, 800 °C 2012 Follow-on to LUNA 1 at higher temperature
LUNA 1 (ICSA)** Inert gas, 500 °C 2012 Evaluation of irradiation damage in fiber optic sensors manufactured by Luna Innovations Inc. for use as in-core temperature monitors
Hydride Fuel* Inert gas, 400 °C (clad) 2011-2012 Irradiation of novel liquid-metal bonded uranium-zirconium hydride fuel rods in Zircaloy-4 cladding for Light Water Reactor (LWR) applications
High-T ICSA irradiation of “Max Phases” ** Inert gas, 300 - 800 °C 2010-2011 Irradiation of five different ternary carbides and nitrides of Ti for post-irradiation evaluation of properties with possible applications in high temperature gas reactors
Mo-99 Irradiation Inert gas (~400 °C) 2010 (Pilot) Evaluate thermal neutron activation (n, γ) production of Mo-99
High Temperature In-core Sample Assembly (ICSA)* Inert gas, to 900 °C 2010 Evaluate thermal design of capsule incorporating “gamma susceptor” for passively heated irradiations up to 900 °C
ACI* Water, 290 °C 2009-2012 Continued irradiation of tubing samples with addition of bonding specimens to evaluate end-cap bond processes
ACI Water, 290 °C 2009 Continued irradiation of clad tubing samples from previous ACI with addition of next generation, smaller diameter tubing
ACI** Water, 290 °C 2006-2007 Investigation of corrosion and mechanical property behavior of triplex SiC/SiC composite clad tubing under PWR conditions
High Temperature Irradiation Facility (HTIF) Inert gas, 800 - 1600 °C 2005-2006 Irradiation of SiC/SiC composites and surrogate TRISO fuel particles
Electrochemical Corrosion Potential (ECP) Water, 290 °C 2004 ECP measurement of in-core coolant and oxides to investigate mechanisms of shadow corrosion
Shadow Corrosion 2 Water, 290 °C 2004 Phase 2 of 1999 Shadow Corrosion Loop
Annular Fuel** Inert gas, 300 °C (clad) 2004 Demonstration of vibro-packed, internally and externally cooled annular fuel under irradiation and test of manufacturing process
Alumina Fiber Composite Clad** Water, 295 °C 2000 Test of alumina-based ceramic fiber composites as potential cladding material under PWR conditions
Shadow Corrosion 1 Water, 290 °C 1999 Test clad samples with various counter materials and gaps under BWR conditions to investigate shadow corrosion
Sensor Irradiation Water, 290 °C 1994 Test in-vessel detectors for Electrochemical Corrosion Potential (ECP) and crack growth rate and investigate crack arrest by hydrogen water chemistry
Irradiation-Assisted Stress Corrosion Cracking (IASCC) Water, 290 °C 1990 Test effects of coolant chemistry on Irradiation Assisted Stress Corrosion Cracking (IASCC) of BWR alloys
BWR Coolant Chemistry Loop Water, 290 °C 1990 Evaluate effect of chemical additives on N-16 carryover and benchmark radiolysis codes for Boiling Water Reactors (BWR)
PWR Coolant Chemistry Loop Water, 320 °C 1989 Measure effect of pH on corrosion product transport and ex-core radionuclide deposition to optimize Pressurized Water Reactor (PWR) chemistry specifications