Research Programs

Advanced Nuclear Power

-- Select program -- A Very Long Cycle Boiling Water Reactor Solid Hydride Fuel for Long-Lived LWR Core Design Investigation of Nanofluids for Nuclear Applications Supercritical Water Reactors Advanced Gas Cooled Modular Pebble Bed Reactors Fission Product Barrier Integrity Fuel Performance Modeling High Performance Alloy Development for Liquid Metal Service Supercritical CO2 Power Conversion System Methods for Risk-Informing Future Reactor Regulation Gas Cooled Fast Reactor Evaluation

Publications:

1. P. Saha, J. Zhao and M. Kazimi, “SCWR Thermal-Hydraulic Instability Analysis” presented at the SCWR Information Exchange Meeting, University of Wisconsin, Madison, April 29-30, 2003.
2. J. ZHAO, P. SAHA, and M.S. KAZIMI, “One Dimensional Thermal-Hydraulic Stability Analysis of Supercritical Fluid Cooled Reactors,” ICONE, April 2004.
3. J. Zhao, P. Saha and M.S. Kazimi, “Analysis of Flow Instabilities in Supercritical Water –Cooled Nuclear Reactors,” MIT-ANP-TR-105 (September 2004).
4. A.J. ZHAO, P. SAHA, and M.S. KAZIMI, “Stability of Supercritical Water-Cooled Reactor During Steady-State and Sliding Pressure Start-Up,” NURETH 11, France, October 2- 6, 2005.

Investigators:

• Mr. J. Zhao
• Dr. P. Saha
• Prof. M. S. Kazimi.

Supercritical Water Reactors

The Supercritical Water-cooled Reactor (SCWR) is one of the six reactor concepts selected in the Gen-IV planning group for further development. The once-through direct cycle version of SCWR combines many positive features of the current generation BWR, PWR and supercritical pressure fossil-fired power plants. In this SCWR design, steam at very high pressure (~25 MPa) and high temperature (500-550oC) enters the turbine, yielding a net plant efficiency greater than 40% which is very attractive for generation of electricity. In addition, the absence of water circulation in the vessel means that the vessel dimensions can be kept small, like a PWR. However, since the temperature rise across the core is about 200ºC rather than 40ºC typical of PWRs, the flow rate can be cut drastically. Both of these features result in lower capital costs. However, the higher pressure requires thicker walls for the primary system and the high temperature requires specific materials to combat corrosion, both of which will increase the capital costs. The net result is expected to be economically favorable.

One of the major technical issues related to the SCWR is the question of possible thermal-hydraulic and nuclear coupled flow instabilities because of the large coolant density change in the reactor core. Both single channel and whole core analyses have been performed using one-dimensional thermal-hydraulics models for the U.S. version of the Gen-IV SCWR reference design. The standard frequency domain approach has been adopted to calculate the decay ratio of the core inlet velocity for a small perturbation in the core pressure drop. Analyses have been performed for both the average and the hot channel of the core. It has been found that the channel inlet restriction (or orificing) plays a major role in maintaining the same power-to-flow ratio among all parallel channels and for avoiding flow oscillation or instability. Sensitivity studies have been performed with respect to the system pressure, reactor thermal power and core flow rate. The trends are very similar to the results reported for BWR flow instabilities, but the sensitivity of the instability margins tends to be larger for the SCWR.

A new map for instability under supercritical conditions was formulated based on a three region representation of the fluid density. Two non-dimensional numbers for the inlet subcooling and the expansion rate in the lightest fluid region were found to control the stability. The whole core regional as well as in-phase oscillations where found to be dependent on the light fluid throttle valve as well as the inlet flow orificing scheme.