Patent Application
20060227924
In the development of nuclear reactors, such as pressurized water reactors and boiling water reactors, fuel designs impose significantly increased demands on all of the core strip and tubular cladding. Such components are conventionally fabricated from the zirconium-based alloys, such as Zircaloy-2, Zircaloy-4, Zirlo, etc. Increased demands on such components are in the form of longer required residence times, thinner structural members and increased power output per area, which cause potential corrosion and/or hydriding problems, as well as issues during design base accident conditions.
Patent descriptions relating to Zircaloy-2 or Zircaloy-4, can be found in U.S. Pat. Nos. 2,772,964 and 3,148,055, (Thomas et al. and Kass et al., respectively). Zircaloy-2 is generally a zirconium alloy having about 1.2-1.7, weight percent (all percents herein are weight percent) tin, 0.07-0.20 percent iron, about 0.05-0.15 percent chromium, and about 0.03-0.08 percent nickel-with the remainder being zirconium. Zircaloy-4 generally contains about 1.2-1.7 percent tin, about 0.18-0.24 percent iron, and about 0.07-0.13 percent chromium.
Other patents in this area include U.S. Pat. Nos. 5,194,101; 5,125,985; 5,112,573; and 4,649,023, (Worcester et al., Foster et al., Foster et al., and Sabol et al., respectively).
While zirconium and other metal alloys have excellent corrosion resistance and mechanical strength in a nuclear reactor environment under normal and accident conditions where the heat fluxes are relatively low, they encounter mechanical stability problems during conditions such as a departure from nucleate boiling (“DNB”) incident that might occur during accidental conditions. Any action tending to increase the heat flux of the core in order to raise the plant output will aggravate these problems.
The response of the fuel to an accident in a nuclear reactor is critical to whether or not a nuclear plant contains the accident and does not release radiation to the environment and people surrounding the reactor. One of the key requirements is that the fuel maintains a coolable geometry. That is, it must be able to keep its basic shape throughout the accident and allow water and steam to pass through the core to maintain temperatures below the point at which it would start to melt and bend into a form which cannot be cooled and heat easily removed from the reactor. In order for this restriction to be met for current metal based fuel clad, grids, and the like, the surface temperature is not allowed to rise above about 1200° C. at any point in the core during a loss of coolant accident. By keeping below this maximum temperature, the fuel rods in a core are guaranteed to maintain their integrity during any design basis accident. The maximum temperature that the clad, grids and the like gets to is a function of the amount of decay heat that is generated and which in turn is related to the power that the reactor operates, and to the rate of reaction between the coolant and fuel rod cladding. The maximum temperature is also a function of the type of heat transfer that occurs between the fuel rod and the surrounding coolant during an accident. In order to balance the cooling heat transfer and the heat generation while still maintaining the required surface temperature, a condition called departure from nucleate boiling (DNB) must be avoided.
Heat transfer modes on the fuel surfaces, for pressurized water reactors, “PWR”, are normally force convection with some amount of sub-cooled nucleated boiling at the upper part of the core (see
As a result of these undesirable conditions, the heat generation rate from nuclear fuel in a PWR is maintained low enough to provide sufficient margins so the occurrence of the departure from nucleate boiling during normal core operation and also during certain reactor transients like Loss of Flow, Locked Rotor or Steam Line Break remains limited or does not occur. Since the portions of the core in the center where new fuel is often added can have much higher heat transfer rates than the rest of the core, the overall power profile of the core is lowered so that this peak critical heat flux is well below that which will cause DNB at the hottest parts of the core. The power density of the core is therefore considerably below that which could be achieved if DNB restrictions were not a design criteria. Additionally, a rather complicated electronic control rod tripping mechanism has been proposed, as described in U.S. Pat. No. 5,631,937 (Robertson). However, this type of active safety approach is not as desirable as the passive approach of restricted heat flux. Similar restrictions are found in boiling water reactor “BWR” fuel where the criterion is to avoid dry-out of the fuel rods. This criterion also limits the reactor operation and is a result of the thermo mechanical properties of the current alloys employed in the cladding.
Obviously there is a need for much higher temperature resistant metals or other materials for cladding, grids, guide tubes, stainless control rods, and other uses in a nuclear reactor environment. One of the main objects of this invention is to provide substitute high heat flux rate resistant materials for use in nuclear reactor environments having the potential for departure from nucleate boiling. Another object is to substantially modify the design criteria and operation constraints by removing or reducing the DNB criterion by using resistant material that allow operations in film boiling conditions (DNB) for limited periods of time without substantive reduction of the mechanical integrity of the fuel rod.
The above needs are fulfilled and the above objects met by providing an article of manufacture for use in the elevated temperature environment of a nuclear reactor, having a possibility of a departure from nucleate boiling, where the article is selected from structural forms within the nuclear reactor environment, such as at least one of cladding for nuclear fuel, support grids, guide tubes and control rods, where the article is coated with or made from a ceramic that is structurally and thermally resistant to temperatures that result from heat fluxes that cause a departure from nucleate boiling. Preferably the ceramic would have a melting point temperature over 1850° C. (about 3362° F.). The most preferred ceramic is SiC.
If the ceramic is to be a coating, over for example a metal alloy, it can be deposited onto the article using plasma spraying or chemical vapor deposition. If the entire article is to be ceramic, the article can be molded, extruded or built up using fibers, solid foam and/or liquid or gaseous precursors of the ceramic. The ceramic used can range from 50% to 100% of theoretical density usually 50% to 95%. The invention also resides in a structural member of any geometry made through any manufacturing technique in a nuclear reactor of any type having a possibility of a DNB, and in a nuclear reactor operating for short periods of time under departure from nucleate boiling having structural forms described above and hereinafter.
The invention as set forth in the claims will become more apparent by reading the following detailed description in conjunction with the accompanying drawing, in which:
The term “nuclear reactor” is meant to include a pressurized water reactor (PWR), a boiling water reactor, a heavy water reactor and the like and any associated auxiliaries, such as turbine generators, fuel cell modules, and the like. The term “departure from nucleate boiling” (“DNB”), besides previous defining descriptions, is also meant to include, where in practice, if the heat flux is increased, the transition from nucleate boiling to film boiling occurs suddenly, and the temperature difference increases rapidly, as shown by the dashed line 24 in
In a PWR, convective heat transfer is used to remove heat from a heat transfer surface. The liquid used for cooling is usually in a subcooled state, at a temperature lower than the normal saturation temperature for the working pressure. Under certain conditions, some type of local boiling can take place on the fuel rods. The most common type of local boiling encountered in nuclear facilities the nucleate boiling. In nucleate boiling, steam bubbles form at the heat transfer surface and then break away and are carried into the main stream of the fluid. Such movement enhances heat transfer because the heat generated at the surface is carried directly into the fluid stream. Once in the main fluid stream, the bubbles collapse if the bulk temperature of the fluid is below the saturation point. This heat transfer process is desirable because the energy created at the heat transfer surface is quickly and efficiently transferred to the bulk fluid.
As local heat flux increases, or due to a modification in the system parameters, such as the pressure/fluid enthalpy flow rate etc., could affect the rate of the creation of the bubbles from the heated surface, no longer assuring that the clad surfaces are continually wetted with liquid water. A transition from nucleate boiling to film boiling occurs and the CHF is reached.
Likewise, if the temperature of the heat transfer surface is increased, more bubbles are created. As the temperature continues to increase, more bubbles are formed than can be efficiently carried away. The bubbles grow and group together, covering small areas of the heat transfer surface with a film of steam. This is known as partial film boiling (18 in
As the area of the heat transfer surface covered with steam increases, the temperature of the heat transfer surface rapidly continues to increase until the affected surface is covered by a stable blanket of steam, preventing contact between the heat transfer surface and the liquid in the center of the flow channel. The condition after the stable steam blanket has formed is referred to as film boiling. The process of going from nucleate boiling to film boiling is graphically represented in
The approach of this invention is to enable the operation of high heat transfer fuel under DNB conditions, for limited periods, by using a clad or other structural material that has a limiting temperature for maintaining mechanical integrity well above the maximum temperature that is generated at the core hot spot during the reference transients. For example, during a locked rotor accident, the heat flux may be greater than the CHF of the DNB criteria and the temperature difference between the fuel surface and the coolant could rise to over 982° C. (1800° F.). This is well above the temperature value for zirconium based clad, where significant weakening occurs. The solution is to use a clad or other structural material/article such as a ceramic, for example, which has a melting point of >2700° C. (4,892° F.), a phase transition temperature of about 1900° C. (3,200° F.) which is well above the highest temperature difference that would give the required heat flux during most of the design basis accidents.
To further understand which articles/components in a nuclear reactor environment would benefit from being made from or coated with a ceramic having a melting point temperature over 1850° C. (such as SiC, the preferred ceramic of this invention); reference is made to
In the embodiment shown in
In most cooled water nuclear reactors, the reactor core is comprised of a large number of elongated fuel assemblies. These fuel assemblies typically include a plurality of fuel rods held in an organized array by a plurality of grids that are spaced axially along the fuel assembly length and are attached to a plurality of elongated thimble tubes of the fuel assembly. The thimble tubes typically receive control rods, plugging devices, or instrumentation therein.
A side view of the fuel assembly is shown in
A liquid moderator/coolant such as water, or water containing lithium and boron, is pumped upwardly through a plurality of flow openings in the lower core plate 64 to the fuel assembly 60. The bottom nozzle 62 of the fuel assembly 60 passes the coolant flow upwardly through the thimble tubes 68 and along the fuel rods of the assembly in order to extract heat generated therein for the production of useful work.
To control the fission process, a number of control rods 72 are reciprocally movable in the thimble tubes 68 located at predetermined positions 56 in
The “structural forms” located within the nuclear reactor environment that are coated with or made from the ceramic of this invention are defined to include at least one of cladding for nuclear fuel, support or mixing grids for clad nuclear fuel, guide tubes (thimble tubes), control rods, lower core support plates, top and bottom nozzles, and instrumentation tubes and the like. Examples of useful ceramics include SiC (melting point >2700° C.); ZrO2 (zirconia, melting point 2700° C.); Al2O3 (alumina, melting point 1900° C.); ZrN (melting point 2930° C.); and AlN (aluminum nitride, melting point 2200° C. at 4 atm and mixtures thereof.
Thus any ceramic type material having a melting point over 1850° C. is useful as the coating or substitute whole article. The preferred materials based on cost/performance are SiC, ZrO2 and ZrN. The most preferred material is SiC. Use of these materials also allows from a 5% to 30% uprate in the power density achievable over metal clad fuel rods without running the risk of the geometrical failure of the fuel during a design basis accident. That is, normal power density for zirconium alloy clad fuel assemblies is about 5 to 10 kw/A. Above this value during a design basis accident, the surface temperature of the clad and perhaps the surrounding grids could exceed the melting point of the clad. However, even at the uprated condition, the clad surface temperature will not exceed the service temperature of the ceramic.
These ceramic type materials are within required radiation, temperature, mechanical and corrosion characteristics required in the nuclear reactor environment. Of course only part of the structural forms need be coated or made entirely of a ceramic type material, for example, the cladding can be coated with or made from ceramic, but the grid can be metal. Thus at least one structural form may contain ceramic and a plurality of other forms may remain metal.
If the previously described materials are to be coated onto a metal surface by coating means such as plasma spraying, chemical vapor deposition or chemical reaction, the thickness should range from 0.01 mm to 10 mm at a density of from 50% to 100% of theoretical density. Over 10 mm and the coating will likely flake off and hinder heat transfer. Under 0.01 mm and there will be insufficient protection of the metal surface from corrosion. Under 50% density and the coating will be too porous for protecting the underlying metal.
If the previously described materials are to be 100% ceramic, that is, for example, all ceramic fuel cladding etc., made by means such as pressing of powders into tubes, winding of tubes from fiber mats, or other forms of the ceramic that have been hardened using a ceramic precursor chemical, then their density should be from 50% to 100% of theoretical density. Under 50% and the tubes will not be gas tight or have sufficient strength.
Having described the presently preferred embodiments, it is to be understood that the invention may be otherwise embodied within the scope of the appended claims.
