In-Core Experiments

In-Core Experiments

In-Core Experiments

The MITR core can accommodate up to three in-core irradiation facilities which have neutron flux levels comparable to a commercial light water reactor; the thermal flux is up to 3.6 x 10 13 n/cm2-s, and the fast flux (>0.1 MeV) is up to and 1.2 x 1014 n/cm2-s. The approximate available dimensions for each in-core facility are ~ 2” diameter x 24” long. The in-core irradiation facilities are suitable for advanced materials and fuel research. (A table of in-core experiment characteristics and constraints is shown below.) The MITR in-core neutron flux level and spectrum are very similar to the levels and spectra of commercial PWRs enabling the real-time testing of in-use and proposed materials and components for power reactors.


Flux graph
Comparison of the neutron spectra for a typical in-core experiment at MITR, and for a typical PWR core. Calculations were performed using MCNP with a three-dimensional MITR model at 5 MW. The MITR in-core neutron spectrum and flux magnitudes are comparable to those of a commercial PWR.


The MITR core is shown below as a diagram and a picture with one experiment loaded. In-core experimental facilities at the MITR are installed by removing a solid dummy fuel element and replacing it with a dummy element that accommodates the desired in-core loop configuration. In some cases the dummy fuel element is integral with the in-core experimental facility structure. Note that these facilities are not permanently installed and are removed from the core tank when they are not in use. This means that a given facility can generally be tailored to meet the requirements of a particular experiment or multiple experiments/specimens.


Core diagram, core with one experiment

Left to right: Core diagram; core with one experiment


On the left is a schematic core map showing the fuel element positions, core structures and control elements. Generally, positions A1, A3 and B4 are occupied by solid dummy elements or in-core experiments. On the right is a photograph of the core with an annular fuel irradiation experiment installed in position A3.

There are two general types of in-core experimental facilities, illustrated schematically in the 3-D model of the reactor shown below. The first, designated “In-Core Sample Assembly” (ICSA), uses an S-bend tube to give access to the in-core irradiation space from the top of the reactor core tank. This facility is generally used for sample irradiations in an inert gas atmosphere with limited requirements for in-core instrumentation (typically thermocouples only). The ICSA is cooled by the reactor coolant, but sample capsules can be insulated to take advantage of nuclear heating for elevated temperature exposures. Active heating or cooling is also potentially available in an ICSA subject to design and safety review. Figure 4 shows the neutron spectra comparison of an ICSA and a typical PWR core.


3D cutaway of reactor core
Cutaway 3-D model of the MITR showing an ICSA (blue) and an in-core water loop (gray) installed in the reactor core (fuel elements removed for clarity).


Another type of in-core facility encompasses a variety of more complex irradiation rigs that are expressly designed for a specific purpose. Past examples include LWR loops used to study various aspects of coolant chemistry, passively and actively loaded mechanical tests under LWR conditions, corrosion test loops for advanced clad materials, a test of internally and externally cooled annular fuel, an irradiation test for high temperature gas reactor materials at temperatures approximately 1000-1600°C, and irradiation of multiple uranium-zirconium hydride fuel rods at PWR temperatures. These experiments are described in more detail in Appendix A to illustrate the types of in-core facilities that can be used and the design envelope that has been previously approved and demonstrated. In some cases, similar experiments can utilize existing irradiation rigs. The modular design and small size of these facilities, however, makes it possible to design and construct new facilities for specific purposes at moderate cost

The MITR also has a license amendment approved by the NRC to perform in-core fuel irradiations as long as the fissile material mass in limited to 100 gm or less, and provided that the fuel irradiation does not contain a forced circulation loop. The MITR is the first research reactor in the U.S. that is licensed to perform in-core fissile materials irradiation.

Members of NRL's In-Core Experiments (ICE) Group have more than 25 years of experience in design and operating in-core experiment facilities. The experiments they can conduct at the MITR include, but are not limited to, the following areas:

  • Corrosion and environmental degradation testing,
  • Nuclear materials irradiations and low dose scoping studies,
  • High-temperature irradiations of cladding and structural materials,
  • Small aggregate fuel irradiations, and
  • Advanced materials and instrumentation development.

A variety of instrumentation can be provided to support in-core experiments, with thermocouple temperature measurement being the most common. Other in-core instrumentation that has been used includes electrochemical corrosion potential and DC potential drop strain and crack growth measurement. Additionally, a wide variety of parameters are routinely monitored and recorded at out-of-core location. For water loops these include temperature, pressure, flow, dissolved hydrogen and oxygen, and conductivity. Real-time residual gas analysis (mass spectrometry) is also available for the ICSA and high-temperature facilities. Radiochemical and chemical assays can be performed on-site on using colorimetry, ICP-OES, and INAA. MITR staff will support use of specialized instrumentation required for a particular experiment, subject to funding and manpower constraints.

 


In-Core Position Characteristics and Constraints

The following table summarizes the general characteristics and constraints of the in-core positions. Note that fueled irradiations can only be performed up to a maximum of 100 g 235U equivalent and that sample fuel must be double encapsulated and cooled by the reactor primary coolant. That is, test fuel cannot be operated in a loop that relies on its own system for circulating coolant.

Parameter Permissible values Comments
Total in-core volume 2” ID x 22” long Maximum available opening in an in-core dummy fuel element
LWR sample space ~ 1” ID x 22” long Typical – dependent on autoclave design pressure and materials.
High temperature sample space ~ 0.8” ID x20” long Dependent on temperature desired and gamma heating susceptor material choice.
MITR coolant wetted materials Aluminum, stainless steel, titanium, zircaloy Small amounts of other materials on case by case basis
MITR coolant heat flux <400 kW/m2 No Onset of Nucleate Boiling
Fissile material content <100 g U-235 or equivalent Fissile materials other than U-235 require pre-approval
Facility reactivity Secured: <1.8% DK/K
Non-secured: <0.5% DK/K
Movable: <0.2% DK/K
“Movable” reactivity limits apply to coolant phase change and dissolvable neutron poison.