Patent Number: 056404344
Section: summary

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a miniaturized nuclear utilizing improved pressure tube structural members. More particularly, the present invention relates to a new miniaturized nuclear reactor utilizing novel structural members that are used to support the loads and stresses of multiple nuclear reactor fuel channel pressure tubes in a confined area. 2. Description of the Prior Art Nuclear power plants traditionally have been designed for achieving long term, safe, and reliable performance. To assure safety, the plants incorporate systems and procedures representing a studied anticipation of emergency conditions. design approaches will have considered theories or premises which may include, for example, design redundancies which are challenged by updated rules of performance as operating experience with nuclear power progresses. Thus, investigators in this power field continuously are called upon to develop improved analytic models of operation exhibiting improved bounding of operational factors and to further achieve higher levels of safety in view of changing rules of safety related performance. Because of the necessarily extensive time interval involved in developing or constructing a new nuclear power facility, for example such an effort may encompass ten years or more, and further in view of the numerous nuclear power facilities now in operation, these investigators typically are called upon to meet new rule criteria by modification of longexisting facilities. Retrofitting procedures can be quite extensive, calling for revised electrical power supplies, major valving replacements, and the like. The nuclear industry has evolved a variety of reactor types. One such type finding substantial field use performs to produce steam for turbine drive within the reactor core itself and is referred to as a boiling water reactor (BWR). The reactor heated water of the BWR serves not only as working fluid, but also as a reaction moderator, and along with other parameters, its proper supply and application within the system necessarily has been the subject of safety requirements or rule generations by government regulatory agencies such as the Nuclear Regulatory Commission (NRC). Typically, the general structure of a BWR nuclear system will include an upstanding reactor vessel which incorporates a lower reactor core structure beneath which are control rod drives. Above the core are, in order, a steam separator assembly and a steam dryer assembly leading to a steam outlet, above the reactor is a shield wall and outwardly of that a drywell. A pressure suppression chamber (wetwell), being torroidal in shape, is located below and encircling the drywell. In more typical BWR installations, water coolant is heated in the reactor core to rise within the reactor vessel as a two-phase mixture of water and steam. This dual phase mixture then passes upwardly through the steam separator assembly and steam dryer structure to enter the steam line leading to a turbine. Following turbine drive, the steam is condensed to water and returned to the reactor by relatively large condensate and feedwater pumps of a feedwater system. The feedwater enters the downcomer region of the reactor, where it is mixed with the water returning from the steam separator and drying functions. The water in the downcomer region is circulated through the reactor core via the vertically oriented recirculation pumps which direct flow to the vertical jet pumps located between the core shroud and vessel wall (downcomer annulus). In typical fashion, two distinct recirculation loops with corresponding recirculation pumps are employed for this recirculation function. In the event of some form of breakage or excursion generating malfunction, referred to as a "loss-of-coolant accident" (LOCA), designers anticipate that the relatively higher temperature-higher pressure water within the reactor will commence to be lost. A variety of safety systems and procedures may then be invoked both for containment and for thermal control of this LOCA. For the latter, thermal control, safety designs recognize that, while loss of the water moderator terminates the core reaction to eliminate a possibility of a nuclear incident, the momentum of generated heat or the residual energy within the reactor will remain of such magnitude as to require a cooling control to avoid for example, core melt down. In general, the amount of water within the containment system is more than adequate for this purpose, for example that contained in the suppression pool, or additionally, the condensate storage tank. To apply this water coolant for the safety purpose, a variety of safety related techniques or "emergency core cooling systems" (ECCS) have been developed to accommodate the LOCA. For example, core spray (CS) systems and low pressure coolant injection (LPCI) installations have been evolved in a variety of configurations. The LPCI system incorporates, for example, four pumps which are activated by a safety system in the event of a coolant loss. Where the loss of coolant is of sufficient extent, and the vessel pressure remains high, for example in the event of a small pipe break then, an automatic safety system will function to depressurize the reactor vessel permitting the relatively lower pressure water supply pumps to operate to introduce water to the reactor. Because the recirculation system earlier described is ideally structured for this purpose, generally it is used by the LPCI system for water introduction under ECCS conditions. Safety designs heretofore have recognized, however, that a recirculation loop may be broken under a LOCA condition. Thus, the pumping of water into that loop under such a LOCA condition may have no effectiveness. Accordingly, the LPCI systems have been equipped with a recirculation loop selection feature termed "loop selection logic" to avoid such conditions. This safety control detects the broken recirculation loop and initiates a procedure injecting water into the redundant, intact recirculation loop by actuating appropriate LPCI injection valves. Experience with such LPCI loop selection features have shown them to be complex and difficult to test and maintain. Under more current rule-based requirements, the design must accommodate for such occurrences as valve failure and the like. However, to function more effectively under current rules, procedures for retrofitting existing facilities to update them are elaborate and quite expensive, implementation involving such activities as recabling, pump reconnection activities and the like. Thus, an approach has been sought by investigators which offers operators the opportunity to eliminate the requirement for a loop selection logic regimen and associated costs therewith while improving the reliability of the LPCI system. Numerous innovations for structural member for nuclear reactor pressure tubes have been provided in the prior an that are described as follows. Even though these innovations may be suitable for the specific individual purposes to which they address, they differ from the present invention as hereinafter contrasted. In U.S. Pat. No. 3,584,903 titled ROLLED CHANNEL JOINTS by inventor James David Prichard, a strong and leak-free hub assembly for use with the pressure tubes of a nuclear reactor is disclosed in which the hub includes a hard insert having at least one groove formed in it, the hardness of the insert being greater than the hardness of the tubular element with which it is joined. Typically, the hub is formed of stainless steel, the insert is formed of surface hardened stainless steel and the tubular element is a zirconium-niobium alloy. The insert has a hardness greater than the hardness of the tubular element. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system. In U.S. Pat. No. 4,555,361 titled METHOD OF REDUCING THE VOLUME OF SOLID RADIOACTIVE WASTE by inventor Leo P. Buckley et at., combustible, solid radioactive waste, such as paper, plastics, rubber, cloth and wood are reduced in volume to ash residue using pyrohydrolysis, a method which combines pyrolysis of the waste in a vessel at temperatures in the range of 500.degree. to 700.degree. C. and gasification of residual carbon with superheated steam. Pressures of 1.0 to 3.5 Mpa are used with steam flows in the range 4 to 50 grams/second/cubic meter so that carbon containing components of the waste are removed as gaseous decomposition products in the form of carbon monoxide and hydrogen leaving an ash residue. The present invention differs from the above described patent due to the features of a method of reducing the volume of ash produced whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance. In U.S. Pat. No. 4,627,069 titled JOULE MELTER FOR THE PROCESSING OF RADIOACTIVE by inventor Keith B. Harvey et at., the joule melter has an outer cylindrical electrode which forms the outer wall of the melt containment, an inner cylindrical electrode which protrudes upward in the containment and forms the outlet for the melt, thus, also determining the depth of the melt. A non-conducting sealing material forms a base plug between the electrodes. A cylindrical electrically conductive baffle is located between the electrodes and includes an opening which allows the melt to flow from near the outer electrode where the melt material is first inserted into the melter, to the inner electrode which is the outlet. In addition to the inner and outer electrodes, the baffle may be connected to a power supply to modify the currents flowing at each of the electrodes. The present invention differs from the above described patent due to the features of melting the radioactive waste whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance. In U.S. Pat. No. 3,837,397 rifled TUBE BUNDLE ASSEMBLY by inventor Michael J. Pettigrew, a robe bundle assembly, for example, a heat exchanger tube bundle or a nuclear fuel element tube bundle, comprises a bundle of laterally spaced tubes, a frame around the outermost tubes, and a lattice of wire cables with their ends held against lateral displacement by the frame and the tubes in the lattice interstices. The cables are deflected round a portion of each tube to space the tubes from one another, and the cables are preferably tensioned against the frame for this purpose. The present invention differs from the above described patent due to the features of the cable matrix whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein. In U.S. Pat. No. 5,213,757 rifled METHOD FOR FIXING A SPRING PACKAGE TO A TOP NOZZLE IN A FUEL ASSEMBLY OF A NUCLEAR POWER REACTOR by inventor Lennart Ohman, a method of fixing a spring package to a top nozzle in a fuel assembly of a nuclear reactor wherein the fuel assembly comprises fuel rods, guide tubes and spacers arranged in a bundle between a top nozzle and a bottom nozzle wherein a T-shaped slot in milled out in a clamp which is welded to or forms an integral part of the top nozzle for receiving one end of the spring package, the end of the spring package is then inserted into the slot and the end is then fixed in the slot by means of a locking pin. The present invention differs from the above described patent due to the features of the bundle whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein. In U.S. Pat. No. 5,213,755 titled LOW PRESSURE COOLANT INJECTION MODIFICATION FOR BOILING WATER REACTORS by inventor David M. Kelly et al., a conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite quantifies from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,788,033 titled CALANDRIA by inventor Luciano Veronesia calandria for use in conducting the hot coolant of a nuclear reactor transversely. The calandria includes an upper plate and a lower plate which support tubes. The plates and tubes are enclosed in a shell which extends above the upper plate and has a supporting flange. The lower plate has holes for transmitting coolant into the region between the plates. The shell has openings whose boundaries mate with the outlet nozzles of the reactor. The tubes are of stainless steel and are dimensioned so that they have mass, stiffness and strength such that they are not subject to failure by the transverse flow of the coolant even at a high velocity. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,788,032 rifled REACTOR WITH FLOW GUIDANCE IN THE UPPER INTERNALS by inventor Jacques Baujat et al., a nuclear reactor has a pressure resistant vertical vessel with inlet and outer pipes situated at the same horizontal level. It also includes internals having a barrel supporting the core and defining with the vessel a down flow path for the coolant from the inlet pipes towards a space under the core and upper internals defining a flow path for the coolant leaving the core, above the latter, and flowing towards the outlet pipes. The upper internals include dividing walls defining circumferentially distributed volumes located at the common level of the pipes and each over a different angular sector. Some volumes belong to the initial part of the down going coolant path and the others force part at least of the coolant leaving the core to follow a path which is successively directed upwardly then curving towards the outlet pipes. The invention is particularly suitable for use in pressurized water reactors. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,759,904 rifled PRESSURIZED WATER REACTOR HAVING IMPROVED CALANDRIA ASSEMBLY by inventor James E. Gillet et al., a calandria assembly is received within the pressure vessel of a nuclear reactor system, at an elevation corresponding to the level of the outlet nozzles of the vessel, and receives pressurized coolant traveling in an axial flow direction within the vessel and turns same to a radial direction for exit though the outlet nozzles. Hollow tubes mounted in parallel relationship at opposite ends to first and second plates of the calandria in conjunction with a cylindrical skin of cylindrical configuration joining the first and second plates of the calandria, present a redundant structure introducing the potential of thermal stresses, which are limited by selection of the pattern of flow holes in the lower plate and the provision of flexible annular weld joints of J-shaped configuration between the lower ends of the calandria tubes and the lower, second calandria plate. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,284,475 rifled WEAR SLEEVE FOR CONTROL ROD GUIDE TUBE by inventor Andrew J. Anthon, a wear sleeve for a guide robe in a nuclear fuel assembly, and a method of installing the sleeve. The sleeve is an elongated metal cylinder having an upper portion adapted to be suspended from the upper end of the guide tube, and a lower portion adapted to be permanently deformed into interference fit with the walls of the guide tube whereby the sleeve may be secured against vertical movement. The method of installing the sleeve includes the steps of suspending the sleeve from the upper end of the guide tube, then expanding a selected lower surface of the sleeve until the sleeve is permanently deformed, whereby an interference fit between the sleeve and robe is formed. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system. Numerous innovations for structural member nuclear reactors utilizing nuclear reactor pressure tubes have been provided in the prior art that are adapted to be used. Even though these innovations may be suitable for the specific individual purposes to which they address, they would not be suitable for the purposes of the present invention as heretofore described. SUMMARY OF THE INVENTION The present invention describes new shapes of fuel tubes. The advantages are that the fuel tubes are stronger and less brittle, there is more surface contact area for heat exchange to take place and therefore, the tubes' new shape is more efficient. For purposes of miniaturization, the fuel tube can be made from any type of material-metal, metal alloys, ceramic, glass, fiberglass, carbon-graphite, epoxy and/or plastic composites or a combination of these materials with or without reinforcements. The surfaces could be enameled, coated, lined and/or cladded. The shapes would be most applicable to the miniaturization, but could be used in the larger scale reactors, USING HIGH ENRICHED URANIUM FUEL. In the present invention, novel fuel bundles for use in miniaturized reactor are described. The novel fuel tube design is to solve partially the problem of disposal of the spent fuel. The secondary benefit is the increased safety of operation of the reactor, in case of accidental meltdown. The present invention describes novel Support pads to hold the fuel tubes in place. The pads are of different shapes and sizes. Pads provide continuous support, intermediate support, are integral with Structural Member and are inserted inside the fuel tube. The pads are applicable to miniaturization. All surfaces and parts of the reactor and/or fuel tubes could be coated, cladded, enameled or lined. The rolled joint connection at the end of the fuel channel pressure tube (FCPT) was developed to facilitate removal and replacement of the fuel channel pressure tube (FCPT). This is important in design of miniaturized reactors and in maintenance of all integral pans contained therein. The advantages of the novel Structural Member Metal Tubes are as follows: 1. PROTECTION OF SURFACE BY LINING OR COATING The interior surface and other integral parts contained therein can be further protected by adding a protective coating or lining of the interior surface of the member metal tube, to prevent irradiation of the metal tube from the FCPT. The coating or lining of the interior surface should be of a material inert to irradiation to provide positive protection. The use of coating, lining, etc. is novel and can be implemented because of the novel design of the novel Structural Metal Member. The deflection and bending stresses inherent therein would be nominal with the present design of the invention. Therefore, the coating or lining would not develop cracks, peel or other structural and/or functional defects. In the prior art, routinely, the calandria tube would deflect and bend to the extent that some coating or lining could not have been used. This obvious disadvantage would be overcome by the present invention. 2. FUEL CHANNEL PRESSURE TUBE (FCPT) The present invention reduces friction for movement of expansion and/or rotation of the fuel channel pressure tubes as follows: A) FCPT made from ceramic or any Irradiation-Inert Material such as glass, fiberglass, carbon-graphite, epoxy, metal alloys, or plastic composites in accord with the following features: 1. Coat exterior surface of the FCPT to reduce friction around the FCPT. PA1 2. Provide steel bends where the FCPT comes in contact with intermediate support pads. PA1 3. Coat the steel bends. PA1 1. Coat exterior surface of the FCPT to reduce friction. PA1 2. Coat to prevent irradiation of support pads and spacers. PA1 3. Coat the ends of the FCPT to prevent irradiation of the tube extension at connection with FCPT. PA1 1. The support pads and separators could be made of metal and/or metal alloys. PA1 2. The metal should be coated at contact with the FCPT to reduce friction. PA1 3. The pads should be grooved or have depressions to allow for circulation and cooling. PA1 1. The support pads and separators should be made out of material ( ceramic, glass, etc.) are inert to irradiation from FCPT. PA1 2. The metal support pads and superstars should be coated or lined with material inert to irradiation from FCPT. PA1 3. The surface of support pad and separators in contact with FCPT should be coated to reduce friction. PA1 4. The pads should be grooved or have depressions to allow for circulation and cooling. PA1 A) providing first and second water flow paths from the source of water coolant to respective first and second recirculation loops; PA1 B) providing low pressure coolant injection pumps actuable or pumping water from the source through the first and second water flow paths; PA1 C) providing a valve arrangement actuable from a closed to an open condition for effecting flow within the first and second water flow path actuating the valve arrangement in response to the safety output to permit water coolant flow simultaneously in each first and second water flow path; actuating the low pressure coolant injection pumps in response to the safety output; and PA1 D) restricting the flow of the water coolant in each first and second water flow path to a predetermined fluid flow rate selected to deliver the predetermined quantity of water coolant to each respective first and second independent recirculation loops, said flow rate being selected as effective for carrying out the emergency cooling of the reactor core from one water flow path. B) FCPT Made From Metal Subject to Irradiation in accord with the following features: 3. NEW SUPPORT PADS AND SPACERS The support pads inserts are one-piece made full length (20) feet of FCPT to be inserted into the new structural metal member inside the tubes. The support pads once inserted to be fastened to the new structural metal member. The fastener(s) should prevent the sliding of the pad out of position. The spacers could be intermittently spaced and do not have to be the full length. They would be held in position by being attached to the full length support pad or attached to the new structural metal member. The support pads could be an integral part of the new structural metal member. The configuration where the web of member penetrates to the inside of the tube as depicted in the drawings. The part projecting part inside the tube to be shaped as support pad or as spacer depending on location. 4. SUPPORT PADS AND SEPARATOR FOR FCPT A) FCPT Made of Ceramic, etc. B) FCPT Made of Metal C) FUEL BUNDLES-SUPPORT PADS INSIDE THE FCPT The fuel bundles inside the FCPT rest directly on the bottom of the robe. The fuel bundles should rest on pads to protect the surface of the FCPT from abrasion, wear and tear. The abrasion is caused by the fuel bundles sliding during loading and unloading and due to the elongation of the FCPT and vibration, etc. The pads would have a shape of rails (two) full length of the FCPT secured at ends against moving out of position. The pads should be used with the FCPT made from metal, ceramic, glass, etc. The pads would also protect the ceramic and glass FCPT from Chipping and cracking. 5. FCPT MADE OUT OF GLASS Advantages of using glass for making the fuel channel pressure tubes. The use would increase safety and reduce radiation emission in case of a meltdown. The glass, during the extreme heat due to meltdown, would melt. The melted glass would encapsulate the fuel bundles and pellets. This would reduce radiation emission from the nuclear fuel and contamination of parts of the reactor. It would minimize the damage to the FCPT effected by meltdown and allow for repairs of the reactor by replacement of the FCPT affected by the meltdown. 6. THE NOVEL REACTOR UNIT The new reactor unit would house four, six, or eight FCPT within it, and be used as a reactor. The unit will be a self contained miniature reactor. The exterior shape is the reactor can be round square, rectangular triangular and polygonal and/or any combination thereof. The tubes as shown in FIG. 8 are novel calandria tubes resting on support pads. Inside the calandria tubes are fuel channel pressure tubes. The use of glass for FCPT would have the advantage of safety, and reduction of emission of radiation during a melt down. In case of meltdown, the metal reactor would be encased to prevent radiation passing to the exterior and placed inside a concrete vault similar to a transformer vault in case of malfunction and/or meltdown the radiation will be contained therein. When the fuel is used up (spent) it will be removed and replaced to provide continuous service of the miniature reactor during normal usage. The present invention of the novel nuclear reactor has support pads for Calandria Tubes. The support pads as shown in FIG. 7 could be used to support the calandria tubes. The load of the fuel bundles inside the FCPT would be transferred to the Calandria tubes and from the Calandria tubes to the new unit reactor. In addition, the pads could be in the shape of two rails on which the bundles could readily slide. The present invention describes novel spent fuel disposal. The fuel pellets of spent fuel could be encapsulated in melted glass for disposal. This could be done individually or in bundles. The glass encapsulated fuel would be encased in concrete blocks to be stacked up in storage. The blocks would be made from contaminated (material) concrete, ceramic, and recast into blocks. The most radioactive is the fuel having a protective shield from a low contaminated material made in shapes for easy shipping, handling and storage. It would be fully automated, requiring no handling by humans. The orderly fashion of disposal would require less space, be economical and would not represent danger to the surrounding area. When the fuel is spent, the fuel bundles are removed from the reactor. The spent pellets should be removed from the bundles. The pellets should intentionally undergo a meltdown, and in the process, some contaminant be added to prevent reprocessing the spent uranium into a bomb grade material (national security reasons) and the pellets should be encapsulated with a glass coating to reduce radiation emission. The pellets should be placed in a storage container. This container should be manufactured from radiation contaminated material. The container could be of metal and/or concrete. The size that could be handles for transport and to put on shelf or warehouse. The process described could be fatty automated and done by remote control. The advantages are that the radiation contaminated material would be utilized and the waste disposal will be done in an orderly and controlled manner. It would reduce the amount of waste, reduce the space to store, and would reduce the amount of radiation from the spent fuel. The controlled and orderly manner of handling and storage would increase safety and protect the environment. The present invention describes disposal of spent fuel being encased in melted down glass. Could be used to dispose of nuclear wastes. The product would be radioactive glass blocks that would have to be stored for safety. The glass blocks would be stable and would not be radioactive molecules leaking. This would be stable for a very, very long time. The present invention describes a novel, state of the art fuel bundle. The fuel bundle is approximately 20 inches long and 4 5/8 inches in diameter. The Cylindrical fuel pellets are approximately 3/4 inches long and 1/2 inch in diameter. The fuel element is a metal fuel. In the present invention, the pads are of different shapes and sizes. Pads provide continuous support, intermediate support, are integral with Structural Member and are inserted inside the tube. The pads are applicable to miniaturization. Another feature of the present invention, is all surfaces and parts could be coated, cladded, enameled, or lined as stated in the text of the first patent and this application. An additional feature of the present invention is the rolled joint connection at the end of the fuel channel pressure tube was developed to facilitate the removal and replacement of the fuel channel pressure tube. This is important in design of miniaturized reactors and in maintenance of all other sizes of reactors, and applies to the use of the fuel channel pressure tube made of all materials. The spent fuel after second use would be less radioactive. It would pose a lesser problem of storage and handling. One benefit of the present invention is in making use of a currently discarded material namely spent fuel in highly radioactive state. An additional benefit of the present invention is in reduction of storage volume of highly radioactive spent fuel. The spent fuel should be used as fuel for "heating" the hot water produced by the second use of fuel in the unminiaturized reactor, would be passed through a heat exchanger and returned to the reactor. The heated water from the heat exchanger could be used to heat apartment and/or office buildings and/or generate electricity and/or generate heat for green houses to produce. The heated water could be convened to steam and utilized as mechanical energy. The spent fuel after first use is still highly radioactive, but not sufficient for production of electricity. The spent fuel when used the second time is less radioactive, and the reactor would also operate at a lower pressure. Still another feature of the present invention is addressed to a structural member for nuclear reactor pressure tubes and method which provides effective insertion of water coolant within the recirculating loops of conventional boiling water reactors, but without resorting to complex loop selection logic. Through analysis by modeling and the like of the requirements of the a structural member for nuclear reactor pressure tubes in terms of time for complete coolant injection and in terms of the required quantity of injected fluid, flow rates of injection are derived and requisite quantities of coolant are determined and identified such that the a structural member for nuclear reactor pressure tubes process is controlled through the simple approach of utilizing flow rate controlling hydraulic resistance within coolant injection conduits. Those hydraulic resistances may be implemented with a conventional orifice, the size and shape of which determines desired flow rates or by the throttling of a valve within the injection conduit achieving the equivalent result. Under the process, cross tie conduits and associated cross tie valving otherwise used for recirculation loop selection for coolant injection are not activated, but merely remain in an open condition under the new method and system, necessary a structural member for nuclear reactor pressure tubes modifications are achieved without resort to the complicated system and instrumentation otherwise required for loop selection with a minimum of hardware perturbation, rewiring or repiping. As another feature, the invention provides a structural member for nuclear reactor pressure tubes having a low pressure coolant injection system for a nuclear power facility of a variety having a boiling water reactor, having a reactor core and normal operating pressure, first and second recirculation loops including respective first and second recirculation pumps and actuable discharge valves, a suppression pool water source, a condensate storage tank, and a safety system responsive to a loss-of-coolant accident to generate a safety output. The system includes first and second low pressure coolant injection pumps having suction inputs and discharge outputs and actuable to pump water. A supply conduit arrangement is provided for coupling the suction inputs of the first and second low pressure coolant injection pumps in fluid flow communication with the suppression pool. First and second coolant injection conduits are provided which are coupled with respective discharge outputs of the first and second low pressure coolant injection pumps and to respective first and second recirculation loops. First and second hydraulic resistance components within respective first and second coolant injection conduits are provided for restricting the flow of water coolant therein to a predetermined fluid rate selected to deliver a predetermined quantity of water coolant to each of the first and second recirculation loops, the flow rates being selected as effective for carrying out the emergency cooling of the reactor core from one coolant injection conduit. A control arrangement is provided which is responsive to the safety output for actuating the first and second low pressure coolant injection pumps. As another feature, the invention provides a method for injecting low pressure cooling water into the boiling water reactor of a nuclear power facility having a source of emergency core cooling water, first and second independent recirculation loops normally circulating water through the core of the reactor for steam generation and a safety system responsive to a loss-of-coolant accident to generate a safety output for effecting the supply of at least a predetermined quantity of water coolant to the reactor, comprising the steps of: As another feature, the invention provides a low pressure coolant injection system for a nuclear power facility of a variety having a boiling water reactor with a reactor core, and normal operating pressure, first and second recirculation loops including respective first and second recirculation pumps and actuable discharge valves, a suppression pool water source, a condensate storage tank, and a safety system responsive to a loss-of-coolant accident to generate a safety output. The system includes first and second low pressure coolant injection pumps having suction inputs and discharge outputs and actuable to pump water. A supply conduit arrangement is provided for coupling the suction inputs of the first and second low pressure coolant injection pumps in fluid flow communication with the suppression pool and further includes a cross fie conduit arrangement for selectively interconnecting the discharge outputs of the first and second low pressure coolant injection pumps. First and second coolant injection conduits are provided which are coupled with respective discharge outputs of the first and second low pressure coolant injection pumps and to respective first and second recirculation loops. First and second low pressure coolant injection valves are provided within respective first and second coolant injection conduits and are actuable between closed and open orientations. Further provided are first and second hydraulic resistance devices within respective first and second coolant injection conduits for restricting the flow of water coolant therein to a predetermined fluid rate selected to deliver a predetermined quantity of water to each of the first and second recirculation loops, the flow rate being selected as effective for carrying out the emergency cooling of the reactor core from one coolant injection conduit. A cross tie valve arrangement is provided within the cross tie conduit which is actuable between open and closed conditions for selectively directing the outputs of the first and second low pressure coolant injection pumps to one of the first and second recirculation loops through select first and second coolant injection conduits. A control arrangement is provided which is responsive to the safety output for actuating the first and second low pressure coolant injection pumps, the first and second low pressure coolant injection valves and retaining the cross tie arrangement in the open condition in the presence of the safety output. The invention, accordingly, comprises the system and method possessing the construction, combination of elements, arrangement of parts and steps which are exemplified in the following description. Accordingly, it is an object of the present invention to provide a new structural member with metal fuel channel pressure tubes that reduce moment, reaction and deflection stresses at the ends of the metal pressure tubes. More particularly, it is an object of the present invention to provide a new structural member that will reduce the incidence of cracks developing in the metal of the fuel channel pressure tubes. The new structural members with ceramic fuel channel pressure tubes reduces moment, reaction and deflection stresses at the end of the ceramic pressure tube. The ceramic pressure tube is not affected by irradiation and growth of its diameter as the metal tube is. In keeping with these objects, and with others which will become apparent hereinafter, one feature of the present invention resides, briefly stated, in the ability to use ceramics instead of metal as the pressure tubes. When the structural member for nuclear reactor pressure tubes is designed in accordance with the present invention, stress of the pressure tube is greatly reduced, if not eliminated. In accordance with another feature of the present invention, the invention provides for the use of ceramic pressure tubes by providing full length support without deflection for ceramic brittle material. Another feature of the present invention is that the new structural member would be made to house four, six or eight, etc. pressure tubes within it. The new structural member would act as a Calandria for all the pressure tubes within. The advantage would be that the new structural member would act as a unit that would nave its own controls as to the flow of gas or heavy water. It could be taken out of service for maintenance or pressure tube replacement, while the reactor would remain in operation. Yet another feature of the present invention is the support pads which cradle the pressure tubes and prevent sideways movement of the tube. Accordingly, it is a general object of the present invention to provide the reduction of stresses in Calandria and pressure tubes. It is a more particular object of the present invention to provide continuous and intermittent support for the pressure tubes. An object of the present invention is to provide the prevention of cracks in the pressure tubes. A further object of the present invention is to eliminate deflection and sag in Calandria and pressure tubes. A still further object of the invention is to provide the use of materials for pressure tubes that withstand irradiation, high temperatures, etc (ceramic). A further object of the present invention is to allow for replacement of pressure tubes without shutting down the reactor. Accordingly, it is an object of the present invention to provide the End Plates of glass or metal will be formed with depressions to fit and accept the ends of the Fuel Elements. More particularly, it is an object of the present invention to provide a Hollow tube to be placed between the End Plates for the length of the fuel bundle. A rod or wire will be threaded through the tube and through a hole in the end plates. After the Fuel Elements will be in place in the End Plates the rod or wire will be reissued and anchored to hold the bundle together. In keeping with these objects, and with others which will become apparent hereinafter, one feature of the present invention resides, briefly stated, in the end of the tube or rod there will be a spacer plate. The spacer Plate will be in contact with the inside face of the End Plates. The stress of the rod or wire will hold the bundle together, but it will not put stress on the Glass Fuel Elements. When the fuel bundle is designed in accordance with the present invention, after the bundle is removed from the reactor, the end plates can be separated from the bundle and reused. The Fuel Elements with the spent fuel could be removed and sent to storage. Still another feature of the present invention is that The End Plate, made of glass or metal will be formed to leave cup-like indentions to fit to accept the ends of the Fuel Elements. Yet still another feature of the present invention is that The End Plate, made of glass, will be (welded) attached to the Fuel Elements by molten glass. Still yet another feature of the present invention is that the End Plate holding the fuel elements together will also be made of glass. Another feature of the present invention is that the Fuel Elements will be assembled into a bundle. Yet another feature of the present invention is that the Fuel Elements will be made of glass and filled with pellets. Still another feature of the present invention is that At Each end, a plate is welded to the Fuel Elements, holding them together as a bundle. Yet still another feature of the present invention is that Approximately Thirty-Seven of the Fuel Elements form a cylindrical Fuel Bundle. Still yet another feature of the present invention is that the Fuel Pellets are stocked end to end inside the cylindrical Fuel Element container and sealed. Another feature of the present invention is that the Fuel element is a metal Fuel Sheathing, a cylinder of approximately twenty inch length and 5/8 inch diameter. Yet another feature of the present invention is that Cylindrical Fuel pellets approximately 3/4 inches long and 1/2 inch in diameter. Still another feature of the present invention is that The Fuel Bundle is approximately twenty inches long and 4 5/8 inches in diameter. The novel features which are considered characteristic for the invention are set forth in the appended claims. The invention itself, however, both as to its construction and its method of operation, together with additional objects and advantages thereof, will be best understood from the following description of the specific embodiments when read and understood in connection with the accompanying drawing. BRIEF LIST OF REFERENCE NUMERALS UTILIZED IN THE DRAWING 10--miniaturized nuclear reactor utilizing improved pressure robe structural members 10 PA0 12--calandria tube 12 PA0 12A--calandria robe coating 12A PA0 12B--calandria robe lining 12B PA0 12C--calandria robe cladding 12C PA0 14--fuel channel pressure tube 14 PA0 14A--fuel channel pressure tube coating 14A PA0 14B--fuel channel pressure robe lining 14B PA0 14C--fuel channel pressure tube cladding 14C PA0 16--fuel bundle support pad 16 PA0 16A--fuel bundle support pad spacer 16A PA0 16B--fuel bundle support pad strap 16B PA0 17--fuel compartment pressure robe 17 PA0 18--fuel channel pressure tube pad 18 PA0 18A--fuel channel pressure tube pad vertical spacer 18A PA0 18B--fuel channel pressure tube pad end 18B PA0 18C--fuel channel pressure robe pad horizontal spacer 18C PA0 20--moderator 20 PA0 22--horizontal interior support pad 22 PA0 22A--horizontal interior support pad proximal end 22A PA0 22B--horizontal interior support pad distal end 22B PA0 22C--horizontal interior support pad groove 22C PA0 22D--horizontal interior support pad concave 22D PA0 22E--horizontal interior support pad coating 22E PA0 22F--horizontal interior support pad lining 22F PA0 22G--horizontal interior support pad cladding 22G PA0 24--vertical support pad 24 PA0 24A--vertical support pad proximal end 24A PA0 24B--vertical support pad distal end 24B PA0 24C--vertical support pad grove 24C PA0 24D--vertical support pad concave 24D PA0 26--fuel bundle 26 PA0 28--angular support pad 28 PA0 28A--angular support pad top member 28A PA0 28B--angular support pad bottom member 28B PA0 30--horizontal exterior support pad 30 PA0 30A--horizontal exterior support pad end 30A PA0 30B--horizontal exterior support pad fastener 30B PA0 30C--horizontal exterior support pad concave 30C PA0 40--fuel bundle 40 PA0 40AA--first fuel bundle proximal end plate 40AA PA0 40AAA--first fuel bundle proximal end plate fuel element end fastener 40AAA PA0 40AAB--first fuel bundle proximal end plate port 40AAB PA0 40AAC--first fuel bundle proximal end plate indent 40AAC PA0 40AAD--first fuel bundle proximal end plate opening 40AAD PA0 40BA--second fuel bundle distal end plate 40BA PA0 40BAA--second fuel bundle distal end plate fuel element end fastener 40BAA PA0 40BAB--second fuel bundle distal end plate port 40BAB PA0 40BAC--second fuel bundle distal end plate indent 40BAC PA0 40BAD--second fuel bundle distal end plate opening 40BAD PA0 40C--fuel element 40C PA0 40D--fuel bundle support 40D PA0 40DA--fuel bundle support proximal end 40DA PA0 40DB--fuel bundle support proximal end spacer 40DB PA0 40DC--fuel bundle support distal end 40DC PA0 40DD--fuel bundle support distal end spacer 40DD PA0 40DE--fuel bundle support spacer tube 40DE PA0 40DF--fuel bundle support rod 40DF PA0 40DG--fuel bundle support nut 40DG PA0 42--reactor wall 42 PA0 44--reactor wall interior horizontal 44 PA0 46--reactor wall interior vertical 46 PA0 48--joint connector 48 PA0 48A--joint connector 48A PA0 50--joiner ring 50 PA0 50A--joiner ring thread 50 PA0 52--service tube 52 PA0 112--second calandria tube 112 PA0 112A--second calandria tube compartments 112A PA0 113--second fuel channel pressure tube support pad 113 PA0 113A--second fuel channel pressure tube support pad end 113A PA0 113B--second fuel channel pressure tube support pad spacer 113B PA0 113C--second fuel channel pressure tube support pad concave 113C PA0 113D--second fuel channel pressure tube support pad convex 113D PA0 113E--second fuel channel pressure tube support pad groove 113E PA0 113F--second fuel channel pressure tube support pad opening 113F PA0 114--second fuel channel pressure tube 114 PA0 114A--second fuel channel pressure tube compartment 114A SECOND EMBODIMENT