WO2017116379A1 - Système de refroidissement d'urgence à fiabilité améliorée pour réacteurs à eau légère - Google Patents

Système de refroidissement d'urgence à fiabilité améliorée pour réacteurs à eau légère Download PDF

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Publication number
WO2017116379A1
WO2017116379A1 PCT/US2015/000454 US2015000454W WO2017116379A1 WO 2017116379 A1 WO2017116379 A1 WO 2017116379A1 US 2015000454 W US2015000454 W US 2015000454W WO 2017116379 A1 WO2017116379 A1 WO 2017116379A1
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WO
WIPO (PCT)
Prior art keywords
nozzle
flow
steam
heat
pressure
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Application number
PCT/US2015/000454
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English (en)
Inventor
Arnold Otto Winfried Reinsch
Original Assignee
Arnold Otto Winfried Reinsch
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Application filed by Arnold Otto Winfried Reinsch filed Critical Arnold Otto Winfried Reinsch
Priority to PCT/US2015/000454 priority Critical patent/WO2017116379A1/fr
Publication of WO2017116379A1 publication Critical patent/WO2017116379A1/fr

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Classifications

    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F04—POSITIVE - DISPLACEMENT MACHINES FOR LIQUIDS; PUMPS FOR LIQUIDS OR ELASTIC FLUIDS
    • F04F—PUMPING OF FLUID BY DIRECT CONTACT OF ANOTHER FLUID OR BY USING INERTIA OF FLUID TO BE PUMPED; SIPHONS
    • F04F5/00—Jet pumps, i.e. devices in which flow is induced by pressure drop caused by velocity of another fluid flow
    • F04F5/14—Jet pumps, i.e. devices in which flow is induced by pressure drop caused by velocity of another fluid flow the inducing fluid being elastic fluid
    • F04F5/24—Jet pumps, i.e. devices in which flow is induced by pressure drop caused by velocity of another fluid flow the inducing fluid being elastic fluid displacing liquids, e.g. containing solids, or liquids and elastic fluids
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F04—POSITIVE - DISPLACEMENT MACHINES FOR LIQUIDS; PUMPS FOR LIQUIDS OR ELASTIC FLUIDS
    • F04F—PUMPING OF FLUID BY DIRECT CONTACT OF ANOTHER FLUID OR BY USING INERTIA OF FLUID TO BE PUMPED; SIPHONS
    • F04F5/00—Jet pumps, i.e. devices in which flow is induced by pressure drop caused by velocity of another fluid flow
    • F04F5/44—Component parts, details, or accessories not provided for in, or of interest apart from, groups F04F5/02 - F04F5/42
    • F04F5/46—Arrangements of nozzles
    • F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
    • F04—POSITIVE - DISPLACEMENT MACHINES FOR LIQUIDS; PUMPS FOR LIQUIDS OR ELASTIC FLUIDS
    • F04F—PUMPING OF FLUID BY DIRECT CONTACT OF ANOTHER FLUID OR BY USING INERTIA OF FLUID TO BE PUMPED; SIPHONS
    • F04F5/00—Jet pumps, i.e. devices in which flow is induced by pressure drop caused by velocity of another fluid flow
    • F04F5/44—Component parts, details, or accessories not provided for in, or of interest apart from, groups F04F5/02 - F04F5/42
    • F04F5/46—Arrangements of nozzles
    • F04F5/469—Arrangements of nozzles for steam engines

Definitions

  • the present invention relates generally to nuclear power plant safety systems, and more particularly to Systems for removing the decay heat in the event of an emergency shut down of the fission process by inserting to graphite rods.
  • the present invention of a passive emergency cooling system using dynamic natural convection is focused on a new concept to improve the reliability of emergency cooling for the current generation of nuclear power plants.
  • Heat removal from a nuclear plant after reactor shutdown is essential to keep the nuclear reactor core from overheating because of the decay heat that is continually generated in the reactor core after the nuclear chain reaction is stopped.
  • the heat removal capability of an emergency cooling system for nuclear reactors alone is not indicative for the probability that it prevents an accident.
  • the reliability of the system and its components is the essential characteristic that determines the probability that the system functions in an emergency. In order to attain the highest possible reliability, all the elements of the system have to be designed for high reliability. Since the failure of a single component of the safety system can lead to the complete loss of its safety function, all parts of the system have to be based on the same physical
  • a safety valve only has to open; then, the safety function of pressure relief by discharging fluid is passive.
  • the control rods of a nuclear reactor only have to drop into the reactor core, the absorption of neutrons that follows and stops the nuclear chain reaction - the safety function proper - is a passive process and requires no continuously moving parts.
  • These types of safety processes cannot be considered passive since one moving part is needed for start-up.
  • the safety function proper of a system using a reactive process is genuinely passive as it does not require continuously moving parts.
  • the present invention uses reactive processes for operation and startup obviating the need for a power supply or continuously moving parts in order to maximize the reliability of core cooling. Consequently, the system has the potential to improve the safety and reliability of nuclear power plants, especially the current generation of nuclear plants, significantly.
  • reactive processes are applied not only to the safety function itself (heat removal) but also to the startup process of the system. If reactive processes were not used in both phases (startup and operation), the overall safety function would be less reliable since the entire system is only as reliable as its least reliable subsystem.
  • the reactive processes utilized by the nuclear safety system of the invention include heat conduction, heat transfer, friction between fluid streams at different velocities, mixing of vapor and liquid, a standing shock wave condensing steam, and potential energy (for example, of an elastic spring or of pressurized gas) for a one-way movement opening the startup valve.
  • the instant emergency cooling system combines mechanical and thermodynamic effects to remove the decay heat from a nuclear reactor and reject it to a heat sink using only reactive processes. This obviates the need for a power supply (e.g. electric or hydraulic), continuously moving parts, rotating or reciprocating pumps, turbines, electric motors and control systems.
  • the reactive system creates a thermodynamic equilibrium at low pressure and low temperature in a mixing tube located down-stream of the nozzles that discharge steam and cooling water at high pressure.
  • the remaining sections of the cooling system steam generator, heat exchanger and connecting pipes
  • the exemplary system is activated by removing electrical power from a solenoid controlled valve which, in the exemplary releases stored energy in a spring to open a startup valve.
  • the flow ratio of steam to cooling water must be selected by the designer so that the saturation temperature and pressure of the mixture is low enough to maintain a vacuum relative to the entire cooling system. As a consequence, steam and cooling water are sucked through the nozzles into the mixing tube. Active components are not necessary to overcome the pressure rise from the mixture at low pressure flowing into the diffusor 21 and heat exchanger 22 at much higher pressure.
  • the system creates two-phase flow of steam and water that passes at supersonic velocity through the low-pressure mixing tube and condenses completely in a shock wave that is capable of overcoming the pressure difference between the high system pressure and the low pressure in the mixing tube.
  • thermodynamic equilibrium created by mixing of fluids at different enthalpies sucked into the vacuum from upstream and discharged by the pressure rise of the compression shock downstream is capable of maintaining the low pressure in the mixing tube.
  • the rest of the cooling system remains at full system pressure which is almost twice that of the pressure in the mixing tube. No mechanical devices such as rotating or reciprocating pumps, compressors or turbines are needed to separate the low- pressure volume from the rest of the cooling system.
  • the thermodynamic process itself maintains the reduced suction pressure in the mixing tube needed for the functioning of the passive coolant recirculation process.
  • This passive condensation and coolant recirculation process creates its own vacuum in the mixing tube by lowering the saturation pressure downstream of the nozzles and, consequently, maintaining coolant flow through the system. It is a form of natural convection based on hydrodynamic forces and thermodynamic phenomena independent from gravity and the typical elevation requirements that apply to regular (hydrostatic) natural convection. As it is generated by hydrodynamic forces, it is referred to as dynamic natural convection.
  • the ratio of water flow to steam flow is an important characteristic of the system.
  • the cross sections of the steam nozzle throat and water nozzle throat that supply fluids to the mixing tube determine this ratio. Although a much larger mass flow of water than steam is required for complete condensation of the steam, the size of the cross sections are similar since the density of water is much higher than that of steam yet the exit velocity from the water nozzle is small compared to the sonic velocity of steam.
  • the two nozzles and the mixing tube do not represent a typical jet pump in the conventional sense of the word. Jet pumps require that the motive fluid is injected at a higher pressure than the passive fluid in order to add sufficient momentum to produce a pumping effect.
  • the proposed configuration maintains steam and water at a uniform pressure level prevailing upstream of the nozzles. Therefore, it can be characterized as a system of variable-diameter conduits that generates a vacuum in the mixing tube and coolant circulation by adding heat at high temperature to the steam generator and removing it at lower temperature from the heat exchanger. It produces a self-sustaining thermodynamic process creating a lower pressure in the mixing tube relative to the rest of the cooling loop remaining at a much higher pressure level.
  • An important design feature of the proposed emergency cooling system is that the same principles are applied to the startup process as to the safety function (heat removal) itself in order to maximize safety and reliability: Only reactive processes which increase the entropy without continuously moving parts are used. Electric power, Diesel generators, thermodynamic cycles or external power supply (all designed for high efficiency) cannot be used for startup and operation in order to gain the highest possible reliability of the safety function.
  • FIG. 1 shows a simplified scheme of the emergency cooling system based exclusively on reactive processes
  • FIG. 2 represents a start-up valve for the emergency cooling system with its actuator using a reactive process to attain maximum reliability for valve opening;
  • FIG. 3 illustrates the design of a valve with its seat at the inner wall of the steam nozzle exit in order to enhance the flow during system start-up while the valve is in the process of opening.
  • FIG. 4 shows an arrangement for cooling the inlet of the water nozzle and the cooling water supply pipe in order to prevent heating of the cooling water in the event of leakage past the closed start-up valve during regular power operation of the nuclear power plant (i.e. before startup of the emergency cooling system).
  • FIG. 5 shows an application of reactive processes for pressure relief in the pressurizer of a pressurized water reactor.
  • Steam may be removed from the reactor vessel of a boiling-water reactor (BWR) or the steam generator in the secondary loop of a pressurized-water reactor (PWR) or the steam generator in the tertiary loop of a liquid-metal-cooled fast breeder reactor (LMFBR).
  • BWR boiling-water reactor
  • PWR pressurized-water reactor
  • LMFBR liquid-metal-cooled fast breeder reactor
  • the emergency cooling system removes the steam, condenses it, rejects the heat of condensation to the ultimate heat sink and returns the condensate to the steam generator without using external power or a control system.
  • the ultimate heat sink for example can be any large body of water such as a cooling pond, river, ocean water, and the water in the torus of a BWR or the condensate tank of a typical nuclear power plant.
  • Any heat sink suitable for long-term absorption of the nuclear decay heat can be used so long as the heat sink can absorb the decay heat of the reactor for the time required by the regulatory agency, for example 72 hours.
  • the heat transfer mechanism on the secondary side of the heat exchanger is hydrostatic natural convection induced by the large temperature difference between the hot water discharged from the mixing tube and the cold water stored in the heat sink.
  • FIG. 1 shows the main steam pipe 12 delivering steam from the steam generator through the main steam isolation valve 14 and pipe 15 to the conventional part of a nuclear plant producing electricity during normal power operation.
  • the reactor will be automatically shut down and the main steam isolation valve 14 will be closed.
  • the core still generates decay heat that has to be removed from the reactor to prevent fuel damage and a release of radioactive fission products to the environment.
  • Opening valve 17 diverts steam from the main steam pipe 12 through pipe 16 and convergent nozzle 18 into the mixing tube 20.
  • convergent nozzle should be understood to mean a nozzle that converges that converges more than it diverges.
  • the steam condenses by mixing with water at lower temperature from the convergent water nozzle 19 supplied by heat exchanger 22 through pipe 23. Even after absorbing the heat of condensation, the mixture temperature is low enough to result in a much lower saturation pressure than the system pressure imparted by the steam generator that serves as pressurizer for the emergency cooling system.
  • 18 is shown as being within the nozzle 19, the positions of the nozzles can be reversed.
  • a significantly lower mixing tube pressure than that of the rest of the emergency cooling system is essential for the passive heat removal process to start and to function.
  • a low saturation pressure and temperature is attained by selecting a suitable flow ratio between steam and water in the nozzles. This is accomplished by designing the system with proper area ratios between the three characteristic flow areas of the system:
  • the ratio of the flow area cross section of the water nozzle throat to that of the steam nozzle throat should be larger than 1.1 : 1
  • the ratio of the smallest flow area of the mixing tube to the flow area of the water nozzle should be larger than 1.3 : 1
  • the ratio of the surface area of the heat exchanger to the flow area cross section of the steam nozzle throat should be larger than 12,000 : 1
  • the ratios between these three areas should be larger than the above limits.
  • Pipe 25 includes a vertical section designed to reduce heat transfer between the saturated water of the steam generator and the components 20, 21 and 23 of the emergency cooling system during normal plant operation. Heating of components 20, 21 and 23 would reduce the driving potential of the emergency cooling system during startup.
  • the high reliability of the present emergency cooling system is based on the utilization of reactive processes that depend on hydrodynamic forces, heat transfer, elastic potential energy and friction to induce coolant circulation.
  • hydrodynamic forces do not exist. Therefore, a hydrostatic phenomenon is used to create a vacuum in the mixing tube and start the coolant flow in the emergency cooling system.
  • the entire cooling loop with nozzles 18 and 19, mixing tube 20, diffusor 21, piping 23, 24, 25 and heat exchanger 22 is filled with water.
  • startup valve 17 (FIG. 1, 2 and 3) about 0.3 to 5 meter above the water level in the steam generator, a negative pressure difference (suction) of about 0.03 to 0.5 bar is created across startup valve 17.
  • the emergency system components can be located inside the containment protected by the thick walls of the containment building against tornados, missiles, terrorist attacks or other external threats.
  • the exception is heat exchanger 22 which may be located outside the containment building in an exterior ultimate heat sink 26 exposed to hazards from external events. Since the heat exchanger rejects the decay heat to the ultimate heat sink, it is of importance for the heat removal function of the emergency cooling system and has to have the same level of protection as all other emergency system components.
  • Missile shield 27 (FIG. 1) is included in the design for this purpose, protecting heat exchanger 22 and the heat sink 26 against missiles, explosions and other external events.
  • Startup valve 17 is the only component of the system that has a moving part required for startup of the emergency cooling system 10. Except for valve 17, therefore, the entire emergency cooling system has a fixed geometry. Moving valve disk 31 out of the flow path produces the geometry necessary for dynamic natural convection to develop in the emergency cooling loop and remove the decay heat from the reactor. All other functions of the cooling system including the rejection of heat to the ultimate heat sink are passive and require no moving parts or power supply for startup and operation. Opening valve 17 by moving the valve disk 31 out of the flow path is essential for the safety function of emergency cooling system 10. Therefore, start-up valve 17 and its actuator 32 have to be designed to function with the highest possible reliability.
  • a reactive process as driver for the valve actuator shown in FIG. 2.
  • the potential elastic energy of a spring 34 is used to move the valve disk 31 out of the flow path of valve 17, opening the flow path.
  • the use of the stored potential energy of spring 34 for start-up is one example.
  • Other forms of stored potential energy such as pressurized gas in a tank can be used.
  • a solenoid coil 32 is energized with the movable core 33 in the position shown, keeping the valve closed. As long as the electric current flows through coil 32 from electrical connection 35 to connection 36, the coil forces the movable iron core 33 downward keeping valve disk 31 in the flow path and valve 17 closed. Loss of electric power or a control system signal cutting electric power would cause solenoid coil 32 to be de-energized and release core 33 so that the spring 34 instantly withdraws disk 31 opening valve 17.
  • FIG. 3 shows another improvement of the startup process of the passive emergency cooling system: Streamlined valve part 41 is seated against the inside of the throat 43 of the steam nozzle 18 closing the flow path completely. Upon partial withdrawal of the streamlined valve part 41 to location 42, the flow resistance during valve opening is reduced, increasing the velocity of the water initially discharged by nozzle 18 and improving the startup characteristic significantly: If only a small fraction of the flow path is open, the throat of a convergent nozzle is formed by the stream-lined valve part 41 and the inside 43 of the steam nozzle throat.
  • a high-velocity water jet is injected initially into the mixing tube and sonic flow can develop faster in the converging steam nozzle 18 although the valve is only partially open during the transition from closed to open.
  • more enthalpy of the steam and water is converted into kinetic energy avoiding unnecessary throttling of the expanding fluids, resulting in a faster and smoother startup of the emergency cooling system.
  • FIG. 4 An additional improvement (FIG. 4) of the reliability of the present emergency cooling system can be gained by removing heat from the piping in the vicinity of the start-up valve during normal operation of the nuclear power plant (when valve 17 is closed and the emergency cooling system is shut down).
  • a potential failure mode during start-up of the emergency cooling system could be caused by leakage of steam from pipe 16 through valve 17 into nozzle 18 and pipe 23 before startup. It would heat up the water in pipe 23 upstream of nozzle 19 so that the heated cooling water could not completely absorb the heat of condensation of the steam in the mixing tube 20 during startup. This could jeopardize the startup of the system, resulting in the failure of its emergency safety function.
  • a temperature sensor 53 attached to cooling jacket 50 would transmit an alarm signal to the control room if the cooling water temperature exceeds a limit during normal operation of the nuclear power plant, informing the operators of the problem in order to initiate a repair of the valve before an emergency occurs.
  • a similar effect caused by thermal pollution from the steam generator through pipe 24 and 25 would also be prevented by the cooling jacket.
  • DNC cooling systems improving the safety and reliability of nuclear power plants is the use as pressure relief system for the pressurizer 66 (FIG. 5) of a pressurized water reactor (PWR).
  • PWR pressurized water reactor
  • this function is performed by power-operated relief valves 69 or safety valves connected to the pressurizer steam space 68. These valves are capable of reliably removing excess energy from the pressurizer and protecting it from overpressure during transients.
  • the valves discharge reactor coolant for pressure relief.
  • the DNC cooling system 60 does not cause any coolant loss in the process of heat removal and reducing the pressure in the pressurizer 66 and the reactor coolant system. Condensing the steam and returning the condensate is more effective than steam removal alone because of the low temperature of the condensate (about 200 °C) returned to the pressurizer at a temperature of about 350 °C.
  • a pressurizer operates similar to a small steam generator at high temperature and pressure. Electric heaters keep the temperature in the pressurizer constant by adding heat if temperature and pressure drop below the lower limits. However, the energy added by the electrical heater elements 70 through cables 71 generates only small amounts of steam sufficient to compensate for transient temperature changes or condensation on the pressurizer walls caused by heat losses.
  • the pressurizer is filled with saturated water 67 which is in thermal equilibrium with steam 68 at a pressure of about 160 bar and a temperature of 350 Degree C.
  • the saturated water 67 of the pressurizer connected to the hot leg of the reactor primary system through pipe 72 imparts the saturation pressure of the water to the entire reactor coolant system.
  • Valve 69 connected to the steam space 68 is closed during normal plant operation to maintain the pressure in the reactor coolant system. If a DNC cooling system is connected to valve 69, opening the valve causes steam from the pressurizer to flow through pipe 61 into the mixing volume 62 and condense by direct contact with water at low temperature from line 64. The actuation of valve 69 at the pressurizer would start the DNC process for emergency cooling and steam relief.
  • Pressurizer steam relief by the DNC cooling system offers several advantages over the current designs using relief valves and especially over the new advanced passive reactors:
  • the DNC system would work like a pressurizer relief valve reducing pressure spikes in the reactor coolant system. However, it would not cause any loss of coolant as relief valves do. This is important in the case of an inadvertent valve actuation.
  • the effect of overpressure protection without potential loss of coolant inventory reduces the nuclear risk from the current nuclear reactor generation and enhances especially the safety of advanced passive reactors.
  • Dissipation of energy is synonymous with an increase in entropy.
  • Energy dissipation is also the essential feature of safety systems. Processes for dissipating energy in safety systems that start with concentrating energy by generating work (Diesel generator and steam turbine of current safety systems) have inevitably a lower reliability.
  • the design feature for emergency cooling described above is the direct and simple way to remove the decay heat from nuclear power plants. Its application would have prevented the accidents at Three Mile Island, Chernobyl and Fukushima-Daiichi. In addition, the required capital investment is much smaller and the probability of human errors significantly lower during emergencies compared to those of active safety systems.

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  • Engineering & Computer Science (AREA)
  • Physics & Mathematics (AREA)
  • Fluid Mechanics (AREA)
  • Mechanical Engineering (AREA)
  • General Engineering & Computer Science (AREA)
  • Structure Of Emergency Protection For Nuclear Reactors (AREA)

Abstract

La présente invention concerne un système de refroidissement passif utilisant uniquement des processus réactifs sans pièces mobiles pour initier sa mise en marche et fonctionner, ce système étant conçu pour maximiser la fiabilité de la dissipation de la chaleur de décroissance pour la génération actuelle de centrales nucléaires et pour les réacteurs passifs avancés. Afin de réduire le nombre de mécanismes défaillants, des processus indépendant de toute source d'énergie externe, telle que le réseau électrique ou les générateurs Diesel, sont exclusivement utilisés pour toutes les fonctions de sécurité. Le système utilise uniquement l'énergie susceptible de provoquer un accident pour faire circuler de l'eau de refroidissement à travers le générateur de vapeur afin de dissiper la chaleur de décroissance, ce qui permet de simplifier la conception et de réduire significativement les coûts d'investissement. La chaleur de décroissance produite par le combustible nucléaire après l'arrêt du réacteur induit la circulation du liquide de refroidissement depuis le générateur de vapeur vers le dissipateur thermique final en maintenant le combustible nucléaire aux températures de sécurité et en empêchant la libération des produits de fission radioactifs dans l'environnement.
PCT/US2015/000454 2015-12-28 2015-12-28 Système de refroidissement d'urgence à fiabilité améliorée pour réacteurs à eau légère WO2017116379A1 (fr)

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PCT/US2015/000454 WO2017116379A1 (fr) 2015-12-28 2015-12-28 Système de refroidissement d'urgence à fiabilité améliorée pour réacteurs à eau légère

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Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US4347623A (en) * 1976-05-10 1982-08-31 Reinsch A O Winfried Flash jet coolant pumping system
US4753771A (en) * 1986-02-07 1988-06-28 Westinghouse Electric Corp. Passive safety system for a pressurized water nuclear reactor
US5398267A (en) * 1993-10-12 1995-03-14 Reinsch; Arnold O. W. Passive decay heat removal and internal depressurization system for nuclear reactors
DE19718867A1 (de) * 1997-05-03 1998-11-05 Winfried Dr Reinsch Passives System zur regelbaren Wärmeabfuhr von Kernkraftwerken
US20140219411A1 (en) * 2013-02-06 2014-08-07 Westinghouse Electric Company Llc Alternate passive spent fuel pool cooling systems and methods

Patent Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US4347623A (en) * 1976-05-10 1982-08-31 Reinsch A O Winfried Flash jet coolant pumping system
US4753771A (en) * 1986-02-07 1988-06-28 Westinghouse Electric Corp. Passive safety system for a pressurized water nuclear reactor
US5398267A (en) * 1993-10-12 1995-03-14 Reinsch; Arnold O. W. Passive decay heat removal and internal depressurization system for nuclear reactors
DE19718867A1 (de) * 1997-05-03 1998-11-05 Winfried Dr Reinsch Passives System zur regelbaren Wärmeabfuhr von Kernkraftwerken
US20140219411A1 (en) * 2013-02-06 2014-08-07 Westinghouse Electric Company Llc Alternate passive spent fuel pool cooling systems and methods

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