Advanced Reactor Technology

Advanced Light Water Reactors

1. High-efficiency fuel for light water reactors (LWRs)
2. Stability of advanced BWRs
3. Nanofluids for nuclear applications

1. High-efficiency fuel for light water reactors (LWRs). The use of advanced fuel instead of conventional solid cylindrical fuel in LWRs has been under investigation by Professor Kazimi and Dr. Pavel Hejzlar. A DOE-supported project investigated the ability to raise the core power of pressurized water reactors (PWR) while maintaining or improving thermal margins via adoption of annular fuel with internal cooling. A special issue of Nuclear Technology was devoted to the results of this effort in October 2007.

A new project supported by Korea Atomic Energy Research Institute (KAERI) was started to examine the application of the annular fuel to Korean Reactors. The effort included assessment of the shutdown margin and effects of partial blockage of the internal coolant due to crud. TEPCO supported a scoping analysis of the use of several advanced fuel designs to raise the power in a boiling water reactor (BWR). The project investigated smaller pins of solid geometry, annular fuel, twisted cross-shaped fuel, and the use of uranium hydride instead of uranium oxide as fuel. The three approaches, with some careful design choices, may be able to raise the power up to 30 percent. In addition, hydraulic tests were conducted and showed that the twisted cross-shaped fuel could reduce the core pressure drop below that of the current fuel, because the twisted rods’ contact points along the axial length eliminate the need for grids for mechanical support. The current focus is on determining the design of an experiment to determine the critical power associated with this new fuel design.

Professor Neil Todreas is investigating the use of inverted fuel concept for LWRs. In a typical fuel rodded design, the fuel rods are surrounded by coolant, while in this design it is the fuel that surrounds the coolant. Hence, the inverted fuel concept designation. The inverted core configuration uses hydride fuel in vertically-oriented hexagonal blocks (U-Th-ZrH_1.6 or Pu-Th-U-ZrH_1.6) perforated by coolant channels arranged in a triangular lattice. A cylindrical Zircaloy clad forms the walls of each coolant channel, and a Liquid-Metal (LM) gap separates the clad from the fuel. Each channel is provided with multiple short twisted tape inserts (TT) aimed at enhancement of critical heat flux (CHF) . A hexagonal duct made of Zircaloy or stainless-steel surrounds each fuel prism and a LM gap separates the inner surface of the duct from the outer surface of the fuel. The inverted fuel configuration yields lower fuel temperature than a typical rodded core having the same fuel volume fraction and coolant pressure drop. However, the addition of TTs reduces this temperature difference, by an extent depending on the TT geometric characteristics. Prediction of pressure drop, heat transfer coefficient and CHF at high pressure in presence of short TTs is currently under investigation.  Neutronic and structural analyses as well as the thermal hydraulic ones will determine the extent of possible power density upgrade.

2. Stability of advanced BWRs. Professor Kazimi and his students examined implications of the coolant density change across the core to density wave oscillations in advanced water-cooled reactors. The stability against coupled nuclear and thermal oscillations in a core cooled by natural coolant convection, in symmetric and asymmetric forms, was analyzed. It was concluded that it is possible to design a core, such as the economic simplified boiling water reactor (ESBWR), to be stable by appropriate selection of core height and inlet orificing. This reactor would be more sensitive to changes in power and flow conditions than is typical of current BWRs at their rated conditions. However, at lower powers, where even a current BWR will operate in natural convection, the ESBWR-like reactor has better stability margins. Furthermore, using an expanded model that allows for the possibility of water flashing, the stability at low pressures conditions was shown to be attainable when appropriate steps in pressure ascendance is taken during the start up of such a reactor.

3. Nanofluids for nuclear applications.  The use of nanofluids (colloidal binary systems of water and nanoparticles) has been shown to enhance the boiling critical heat flux (CHF), which can ultimately result in power density increases in boiling systems, such as nuclear reactors, high-power electronics, heat exchangers, etc. In the past year, Professor Buongiorno and Dr. Linwen Hu’s research program on nanofluids has focused on studying the detailed features of nanofluid boiling. The work is enabled by the deployment of a unique pool boiling facility equipped with infra-red thermography, which allows for acquisition of the two-dimensional time-dependent temperature distribution on the boiling surface. Traditional hard-to-measure parameters such as bubble frequency, departure diameter and nucleation site density are accurately obtained with this facility, and can reveal the subtle differences between nanofluid and water boiling behavior. Data interpretation is underway.

An experimental investigation of nanofluid convective heat transfer in the turbulent and laminar flow regimes was completed and revealed that, if the measured properties of the nanofluids are used in defining the governing dimensionless groups, the traditional correlations/models accurately reproduce the heat transfer and viscous pressure loss behavior of the nanofluids. This is a significant finding because controversial abnormalities in nanofluids heat transfer had been reported previously in the literature.

At the first scientific conference entirely dedicated to nanofluids (Nanofluids: Fundamentals and Applications , September 16-20, 2007, Copper Mountain, Colorado), it was decided to launch an international nanofluid property benchmark exercise (INPBE) to validate nanofluid property measurements (particularly of thermal conductivity) performed with various experimental methods, and to generate a reliable database of nanofluid properties. Twenty-four organizations from the US, UK, France, Switzerland, South Korea, India, China and Singapore participate in the exercise, which is coordinated by Prof. Buongiorno. More information on INPBE is available at http://mit.edu/nse/nanofluids/benchmark/index.html.

The nanofluid research program is in collaboration with the MIT Nuclear Reactor Laboratory (NRL) and is sponsored by AREVA, DOE, EPRI, Saudi Arabia’s King Abdulaziz City of Science and Technology (KACST) and a generous gift from Mr. Doug Spreng.

MIT Reactor Upgrade

Use of Low-Enrichment Fuel in the MIT Research Reactor.  The use of highly enriched uranium in research reactors has facilitated high neutron fluxes for use in many scientific disciplines. Last year, Dr. Thomas Newton, working with Professor Mujid Kazimi, Professor Benoit Forget and Dr. Edward Pilat, developed a design of the MIT reactor core using monolithic 20 percent enriched uranium 7 percent molybdenum (U-7Mo) fuel that maintains acceptable thermal and fast neutron fluxes within the confines of the existing core structure by using plate-type fuel with the same dimensions as the current fuel elements. This year, they were engaged in an effort to qualify the methods used for thermal-hydraulic and neutronic safety evaluations of the new core and in examining a code from Argonne National Laboratory to do fuel management studies more quickly.

Increasing the Si Irradiation Capability at MIT Reactor. As part of Professor Kadak’s design course, a conceptual design was developed for increasing the capability of the MITR to irradiate larger silicon ingots. The current capability is limited to 4 and 6 inch ingots with a market demand for larger 8 inch ingots. The students developed an innovative design that used an existing 12 inch beam port in a non-conventional way to allow for the insertion and removal of ingots without requiring a redesign of the reactor support structure. The students additionally reviewed the existing irradiation facility and proposed modifications that would improve throughput and reduce occupational exposure. This project prompted interest by a commercial firm to take the next steps in design to implement such a system.

Enhanced Computational Reactor Physics. Professor Forget is establishing a computational reactor physics group since his arrival in January 2008. He and a graduate student are currently working on the development of a Monte Carlo fuel cycle code that will support the MITR conversion. This code enables fuel shuffling and fuel management capabilities to MCNP/MCODE and will be extended to other reactor types beyond the MITR reactor. He is also looking to establish new methodologies that will increase the fidelity of current simulations of nuclear reactors by improving and transforming current modeling and simulation approach.

Advanced Fast Reactor Systems

A three-year research project is supported by the Nuclear Energy Research Initiative of the Department of Energy. Its objective is to develop risk-informed design development and evaluation tools that take into account safety, economics, licensability, and proliferation resistance. These tools are applied to a number of design alternatives to identify opportunities to reduce the cost of the sodium-cooled fast reactor while maintaining a high level of safety and proliferation resistance. The intent is to assist DOE in its planning purposes, in the development of technical requirements to be imposed on the industrial design organization, in the identification of research needs, and in assessing the technology risk of alternatives.

The project is led by the Massachusetts Institute of Technology and includes The Idaho State and The Ohio State Universities. Professor Apostolakis is the Principal Investigator. Dr. Hejzlar and Professors Driscoll, Golay, Kadak, and Todreas are leading individual tasks and contribute to the overall direction of the project. A Review Group consisting of senior representatives from General Electric Company, and Argonne, Idaho, and Lawrence Berkeley National Laboratories provides guidance and access to relevant information.

Corrosion resistant, functionally graded composite material for Pb-Bi cooled reactors.  Professor Ballinger’s group has been developing a composite structural alloy that will be resistant to corrosion in high temperature liquid Pb and Pb-Bi eutectic to temperatures as high as 700°C. A new corrosion resistant alloy, developed by Professor Ballinger’s group, is being applied as a cladding layer to both the inside diameter of Grade T91 pipe and the outside diameter of Grade T91 fuel cladding. The project is in its’ third year and the final product will be completed and corrosion tested. Separate effect studies to explore the evolution of the interface between the corrosion resistant and structural layers has shown that little dilution will occur over the expected life of either pipe or fuel cladding. Successful completion of the project will result in technology that will enable viability of the Pb-Bi system from a materials standpoint as well as allowing higher temperature operation. Higher temperature operation will greatly improve the overall economics of the Pb-Bi system.

Corrosion of materials in supercritical CO₂. A program is under way in Professor Ballinger’s group to explore the corrosion of materials in supercritical CO₂ over the temperature range 650-800°C and the pressure range 12.5-25 MPa. A wide range of materials are being studied initially. Once initial results are obtained the matrix will be narrowed and the temperature-pressure range expanded as well as the length of exposure. Professor Ballinger’s laboratory (H. H. Uhlig Corrosion Laboratory) is the only university laboratory with this capability. A detailed analysis of the corrosion process is being carried out with a goal of optimizing performance through material and processing/heat treatment control.

Advanced Gas-Cooled Modular Pebble Bed Reactors

The Next Generation Nuclear Plant (NGNP) is being proposed for construction at INL as part of the US effort to demonstrate non-CO₂-emitting methods to produce hydrogen. CANES continues to work on the development of the pebble bed reactor, which is one of the two high-temperature reactors being considered for the NGNP. This year an air ingress experiment was completed by two students in the Aero/Astro department as part of their year long design project under the supervision of Professor Kadak. The students were able to successfully demonstrate that the injection of very small amounts of helium in the top of the reactor vessel upon a break in one of the primary pipes, the onset of natural circulation could be delayed if not completely prevented. If this proves to be valid for larger reactors, this injection system could avoid a major safety challenge of high temperature gas reactors.

An additional experiment completed by one of Professor Kadak’s undergraduate students was a pebble flow experiment to address bypass flow through a central dynamic graphite reflector. The issue is with a large column of pure graphite pebbles in the center of the reactor, a considerable amount of flow that is not heated would bypass the fueled region requiring operation a higher temperatures since the helium coolant is eventually mixed in the lower portion of the reactor. This initial test was to show that if smaller central graphite pebbles were used compared to the outer fuel pebbles, the laminar flow paths previously demonstrated would be maintained with the different size pebbles. The result was that the mixing did not occur allowing for the next flow tests which are being designed by another UROP student.

Progress also continues on modularization of pebble bed reactor designs with Professor Kadak. A student completed a thesis validating that the approach is practical by working with General Dynamics Electric Boat, one of the two builders of the US nuclear submarine fleet. Electric Boat pioneered the modular integrated engineering, design and fabrication of submarines. Additionally, a master’s thesis is currently being completed on optimizing the size of the reactor vessel from a core physics standpoint so that the vessel can be readily shipped by train to locations that are not near coastal regions to allow for a broader application of this technology to support the “lego” construction of this plant. CANES continues collaborations on the Chinese pebble bed project sponsored by the Institute of Nuclear Engineering Technology of Tsinghua University. Professor Kadak participated in an International Atomic Energy Agency meeting in October 2007 on high-temperature reactor safety in Beijing. He also collaborates with the South African PBMR project and the Westinghouse design team for the Next Generation Nuclear Plant.

Material performance in high temperature gas-cooled reactor (HTGR) systems. A consortium of universities has been established with MIT as the technical lead to develop and understanding of the behavior of alloys proposed for use in HTGR heat exchangers that will operate at temperatures as high as 950°C. Professor Ballinger is the MIT PI. The partner universities are the University of Nevada-Las Vegas, Boise State, and the University of Illinois-Urbana. The Idaho National Laboratory is also a participant. The team is exploring the mechanical behavior of alloys 617 and 230 over the temperature range 450-950°C. Properties being evaluated include creep, creep-fatigue, creep crack growth, fatigue and static crack growth in prototypic HTGR environments. Initial results indicate that in some temperature ranges there will be a significant environmental effect when the oxygen potential becomes comparable with the equilibrium potential for the metal-oxide.