Patent Number: 049833537
Section: summary

BACKGROUND OF THE INVENTION This invention relates to steam generators for use in sodium cooled nuclear reactors More particularly, a steam generator is disclosed which has an alternate path for relieving a sodium water reaction and injected water/steam upon a tube casualty causing a violent sodium water reaction STATEMENT OF THE PROBLEM Sodium reactors utilizing sodium contain two discrete sodium loops. A first and primary loop is in the reactor vessel. This primary sodium loop is radioactive and consequently is confined to the reactor vessel. Sodium in this loop circulates through the reactor core in a so-called hot leg and to an intermediate heat exchanger. In the intermediate heat exchanger, the heat from the nuclear reaction from the reactor core is transferred to sodium in a secondary and nonradioactive loop. Once the sodium of the primary loop liberates its heat, it then passes in a so-called cold leg to pumps that force the required circulation. The pumps, preferably interior of the reactor and of the electromagnetic variety, force the sodium of the primary loop through the reactor core so that the cycle of the primary loop endlessly repeats. A second and secondary non-radioactive sodium loop transports the heat of the nuclear reaction from the intermediate heat exchanger in the reactor vessel to a steam generator, the steam generator being located exterior of the reactor vessel. Typically the sodium of the secondary loop passes through an intermediate heat exchanger within the reactor vessel. In this intermediate heat exchanger, it receives the heat from the sodium of the primary loop and returns from the reactor vessel in a so-called hot leg to the steam generator. At the steam generator, the hot sodium is cooled by counterflow heat transfer to water which is heated and turned into steam. This water flows upwardly through helically coiled tubes. The liquid sodium flows downwardly on the outside of the helically coiled tubes, these tubes being immersed in the hot sodium. Steam is generated for conventional electrical power production by turbines, generators and condensers. The sodium of the secondary loop then passes in a cold leg through pumps interior of the steam generator which force circulation. The cooled sodium flows back to the intermediate heat exchanger in the reactor vessel for an endless repetition of the cycle of the secondary loop. A preferred form of a steam generator is set forth in U.S. patent application Ser. No. 231,031, filed Aug. 11, 1988 and is entitled Compact Intermediate Heat Transport System for Sodium Cooled Reactor, now U.S. Pat. No. 4,905,757, issued Mar. 6, 1990. In this steam generator the defined secondary loop includes a sodium surge volume capped by an inert gas, a central concentrically mounted pump, and a steam generator having helically coiled tubes. All of these components are contained interiorly of a single, upstanding substantially cylindrical vessel. The cylindrical vessel has its axis vertically disposed and includes an inner concentric cylindrical vessel open at the bottom to a plenum. The interstitial volume between the inner cylindrical vessel and the main steam generator vessel contains the tube bundle and is used for steam generation at the termination of the hot leg of the secondary loop. Typically four tubes sheets at the bottom of the steam generator and four corresponding tubes sheets at the top of the steam generator serve as the respective beginnings and ends of tubes of the steam generator. These tubes are helically wound in the cylindrical interstitial volume between the outside of the inner cylindrical vessel and the inside of the outer cylindrical vessel which is the steam generator shell. Water to be vaporized into steam flows from the bottom of the generator to the top of the generator. Sodium, for heating the water in the tubes, flows from its inlet nozzle at the top of the cylindrical vessel, through the shell-side of the tube bundle, to the bottom plenum of the cylindrical vessel. At the bottom of the vessel, the sodium passes into a common plenum connecting the outer cylindrical vessel and the inner cylindrical vessel. A pump, preferably of the electromagnetic variety, may be located within the inner vessel. This pump pumps the sodium in the cold leg of the secondary loop upwardly through the interior of the inner vessel to the top of the steam generator. At the top of the steam generator a volume of inert gas accommodates thermal expansion/contraction of the sodium in the system. Discharge of the sodium from the top of the vessel to the main reactor enables the secondary loop to endlessly repeat. Sodium water casualties in the secondary loop of sodium reactors are known. It is, of course, also well known that the reaction between sodium and water is a violent and explosive reaction. It has been found that sodium water reactions in sodium heated steam generators commonly effect more than one tube. Indeed, for the purposes of the sodium water casualty scenario guarded against by this invention, it is assumed that a great many tubes are effected. To relieve pressure caused by sodium water casualties in such steam generators, the prior art has caused a rupture disk to be installed at the plenum on the very bottom of the steam generator. This rupture disk constitutes a mechanical fuse which is broken by the shock of the violent sodium water reaction. Upon the sodium water casualty, this fuse ruptures and permits the steam generator to be emptied of liquid sodium, generated hydrogen, sodium oxide, sodium hydride, water and steam. In order for such emptying to occur, components of the reaction must pass from the site of the ruptured tubes thru the remaining intact tube bundle before the reaction components can pass outside of the steam generator at the ruptured membrane. Assuming that such a casualty occurs, at least three aggravating circumstances can be assumed to be present for the casualty scenario herein addressed. These aggravating assumptions would be required by the National Regulatory Commission (NRC) unless an inherent or passive device, such as the subject of the non-safety related steam and feedwater isolation and blowdown valves. First, it will be assumed that the main steam and feedwater line isolation valves are either inoperative or, if operative, not properly actuated In either case, the site of the ruptured tubes will be supplied with a steady supply of steam and/or feedwater. The steam will react with the sodium in a continuing violent reaction. The high temperatures produced by the reaction can be assumed to increasingly penetrate the secondary loop through an increasing number of ruptured tubes. Second, it will be assumed that the steam/water dump system valves are either inoperative or, if operative, not properly actuated. Therefore, the steam/water inventory in the steam generator system is not vented to the atmosphere, and the steam/water supplied to the site of the ruptured tubes is not diminished by the dump system. Third, it will be assumed that the tube rupture occurs in the singular most undesirable location. This undesirable location is in the upper portion of the steam generator tube bundle. In such a location, the reactive components generated by the violent continuing sodium water reaction will have to pass over the intact lower tubing before escape can occur from the plenum of the steam generator vessel at the ruptured diaphragm. DISCOVERY OF THE PROBLEM We have discovered that it is more than conceivable that with a tube rupture at the top of prior art steam generators, a continuous supply and venting of steam and/or water will generate a pressure differential between the sodium inlet and outlet nozzles of the steam generator. This pressure differential will continue after the steam generator has been relieved of sodium because of steam/water flow friction between the site of the tube ruptures and open disk in the bottom of the steam generator. Simply stated, steam from the site of the broken or ruptured tubes will have to flow past the remaining intact helically coiled tubes to escape. In other words, the site of the initial tube breakage will remain under relatively high pressure in an environment of continuously supplied steam. Assuming that steam is generously and continually supplied to such a casualty site, the broken, secondary sodium loop will experience a continuing and advancing inundation of invading steam. That is, the continuously reacting high temperature Na/steam interface will be forced by the pressure differential within the steam generator to move in a backward direction down the secondary hot leg piping toward the IHX which is submerged in the radioactive primary sodium. Logically, and completing the casualty scenario, the secondary loop will be penetrated by the steam until invasion of the main reactor vessel occurs thru a high temperature failure of thin walled IHX tubes. Specifically, the radioactive sodium of the primary loop could become involved in the sodium water casualty if the casualty were allowed to proceed. It should be understood that the movement of high temperature reacting Na/steam interface towards the primary loop and downward into the IHX will only occur if the continuing steam flow entering the shell side of the steam generator produces a sufficient pressure differential (12 psi or more) as it passes thru the tube bundle to the rupture disk to overcome the elevational difference between the IHTS piping and the IHX. Progression of the Na/water-steam interface into and within the main reactor vessel thru ruptured IHX tubes could conceivably occur. We are unaware of the prior art considering this casualty scenario Insofar as discovery of this problem can constitute invention, invention is therefore claimed SUMMARY OF THE INVENTION In a steam generator utilized with a liquid sodium cooled nuclear reactor, provision is made to vent the violent explosion emanating from tube rupture. Tube ruptures in such steam generators causes a sodium water reaction which in this disclosure is vented along two discrete paths to assure that under no condition will it be possible to develop a sufficient pressure differential to force the Na/steam interface backward and downward into the thin walled IHX tubes In the preferred embodiment, the steam generator is of the type combining a thermal expansion volume, one or more electromagnetic pumps with or without a jet pump to increase the flow rate for circulating sodium, and a large number of helically coiled tubes for generating steam. The steam generator includes a sodium plenum at the bottom thereof and a conventional rupture disk for venting sodium, steam, hydrogen and other reaction compounds immediately upon a tube rupture casualty. The steam generator is contained within a cylindrical vessel and defines interior of the vessel an outside and annular downcoming hot leg and an interior and concentric upcoming cold leg. These hot and cold legs are interconnected at the bottom of the steam generator vessel at the plenum. Typically, the steam generating tubes are commenced at tube sheets at the bottom of the steam generator, are helically coiled around the periphery of the inner concentric vessel interior of the steam generator, and have water channelled therein to counterflow the downcoming liquid sodium in the hot leg of the secondary loop. The casualty scenario anticipated by the disclosed safety feature includes a rupture of a large number of tubes in the top of the steam generator. This rupture is presumed to have a continuous supply of steam and generates back pressure continuously at the site of the tube rupture and accompanying violent reaction It is presumed that the rupture diaphragm opens but the remaining intact tubes present flow resistance to the escaping steam flowing past the remaining intact generator tubes between the rupture site and the open diaphragm. This back pressure forces the steam through the secondary sodium loop to and towards the IHX in the nuclear reactor vessel for eventual invasion into the main reactor with chemical reaction of the radioactive sodium in the primary loop. To prevent this casualty scenario, the invention includes providing an alternate concentric flow path interior of the steam generator. This alternate concentric flow path extends from the upper portion of the steam generator down into the plenum adjacent the diaphragm. This alternate path is filled with sodium during normal reactor operation. Upon a casualty, the alternate flow path dumps its sodium through the conventional rupture disk and then provides an immediate additional pressure relief path for steam, hydrogen and sodium from the site of the violent reaction at the tube rupture site. Consequently, the steam venting pressure drop within the shell at the steam generator can be limited to a value which is insufficient to drive the interface downward into the IHX which behaves like a U-tube manometer. Other Objects, Features and Advantages An object to this invention is to provide a sodium steam generator with a vent path from the top of the steam generator to the plenum at the bottom of the steam generator. This parallel alternate vent path provides for immediate pressure relief of a tube rupture casualty adjacent the top of the generator. Accordingly, and in the preferred embodiment of the invention, a second vent path from the cover gas region at the top of the steam generator is provided down to the plenum at the bottom of the steam generator. Preferably, this second vent path is provided by a second concentric wall surrounding the inner cylindrical vessel. This second concentric wall extends above the sodium region into the inert gas cover at the top of the steam generator. It also extends down from the top into the plenum. During normal operation, this vessel defines an interstitial volume between the inner cylindrical vessel and the secondary vessel. This volume is filled with a standing and normally static head of sodium supported on the low pressure zone in the secondary loop from the reactor plenum. During normal operation, this vent path is filled with sodium from the low pressure region of the secondary loop. During a tube casualty, this tube immediately looses its sodium content by flow to the plenum which is opened at the diaphragm. The conduit empties of sodium and provides a direct flow path from the top of the reactor and out the ruptured diaphragm. Consequently, pressure is relieved and steam inundation of the secondary loop to and towards the nuclear reactor vessel is avoided. A serendipitous advantage of the preferred embodiment is that the sodium in the concentric conduit is drawn down in normal steam generator operation under the low pressure within the steam generated outlet plenum. Consequently, the cover gas region over the top of the secondary conduit is likewise extended downwardly. This extended cover gas region forms a cylindrical annulus between the hot downcoming peripheral sodium in the hot leg of the steam generator and the upcoming pumped sodium in the center portion of the steam generator. Consequently, heat from the hot leg is prevented from shunting across to the sodium of the cold leg. More efficient insulation of the cold and hot legs interior of the steam generator results.