Safety of Nuclear Power Plants Generation I to Generation IV
Transcrição
Safety of Nuclear Power Plants Generation I to Generation IV
Safety of Nuclear Power Plants EPR,, GIF,, HTR Generation I to Generation IV Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 2 1 EPR Overall Safety Philosophy These objectives, motivated by the continuous search for a hi h safety higher f t level, l l involve i l reinforced i f d application li ti off th the defense in depth concept: by improving the preventive measures in order to further reduce the probability of core melt, and by simultaneously incorporating, right from the design stage, measures for limiting the consequences of a severe accident. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 3 EPR Plant Parameter Thermal power 4250/4500 MW Electrical power 1600 MW Efficiency 36% No. of primary loops 4 No of fuel assemblies 241 Burnup > 60 GWd/t Secondary pressure 78 bar Seismic level 0.25 g Service live 60 years Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 4 2 Risk Reduction – EPR Approach Reinforced protection in the unlikely event of core meltdown Preventive features include safetyy devices which further reduce the probability p y of a severe accident enlarged water inventory of the main primary system and of the steam generators; increased reliability of safety systems through 4-fold 100% redundancy (4-train concept); use of diversified technologies for each train of these systems. Features to mitigate the consequences of such an event: the extremely robust, leaktight containment around the reactor is designed to prevent radioactivity from spreading outside; the arrangement of the blockhouses inside the containment and hydrogen catalytic recombiners (passive devices) prevent the accumulation of hydrogen and the risk of deflagration; molten core escaping from the reactor vessel would be passively collected and retained, then cooled in a specific area inside the reactor containment building (“core catcher”) Enhanced protection against external hazards Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 5 EPR – Safety Systems Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 6 3 EPR Safety Injection / Residual Heat Removal (SIS/RHRS) Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 7 Risk Reduction - EPR Containment The unique building housing the reactor is extremely robust. It rests on a 6 m thick concrete base mat and is enclosed by a double shell: the inner is made of leaktight, prestressed concrete and the outer one of reinforced concrete, each 1.30 m thick. This total to 2.60 m concrete thickness, capable of withstanding external hazards such as an aircraft crash. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 8 4 EPR Core Catcher Even in the event of the core melting, and piercing then escaping from the steel reactor vessel in which it is housed, it would be contained in a dedicated spreading compartment. This compartment is then cooled to remove the residual heat. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 9 EPR Containment Heat Removal System Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 10 5 Generation IV International Forum (GIF) To meet future energy needs, ten countries have agreed on a framework f k for f international i t ti l cooperation ti in i research h ffor an advanced generation of nuclear energy systems, known as Generation IV. Euratome joined as 11th full member; Switzerland joined in February of 2002 Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 11 Generation IV – Overall Goals Development of one or more nuclear energy systems Deployable by 2030 With significant advances in: Sustainability Safety and reliability Proliferation and physical protection Economics Competitive in various markets Designed for different applications: Electricity, Hydrogen, Clean water, Heat Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 12 6 Generation IV Initiative – Aims (1/2) Nuclear energy-systems including fuel cycles of the 4th generation ti should h ld ... SR-1 ... be excellent regarding safety and reliability while in operation. SR-2 ... have very low core damage frequency and little consequences. SR-3 ... eliminate the need for emergency g y planning g outside of the plant. EC-1 ... clear lifecycle-cost advantages over other energy sources. EC-2 ... comparable financial risk to other energy projects. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 13 Generation IV Initiative – Aims (2/2) Nuclear energy-systems including fuel cycles of the 4th generation should ... SU-l ... produce sustainable energy, following regulations for air pollution prevention and enhancing the long term availability of the system and efficient usage of fuel. SU-2 ... minimise the nuclear waste and disposing it, especially they g term scale and will reduce administration efforts on a long hence improve the protection of health and environment. SU-3 ... increase the certainty that they are an undesirable and difficult source to obtain dangerous materials for usage in weapons. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 14 7 Selected Generation IV Concepts (out of 21) GEN IV Concepts Acronym Pressure Spectrum Fuel cycle Temperature Sodium Cooled Fast Lead Alloy-Cooled Gas-Cooled Fast Very High Temperature Supercritical Water Cooled Molten Salt SFR LFR GFR VHTR SCWR MSR Fast Fast Fast Thermal Th.&Fast Thermal [°C] 530-550 550-800 490-850 640-1000 280-510 565-850 Spring 2011 / Prof. W. Kröger Closed Closed Closed Once-Through Once/Closed Closed [bar] 1 1 90 70 250 <5 Safety of Nuclear Power Plants 15 The Roadmap Addresses Viability and Performance R&D Phases Viability Key K ffeasibility ibilit and d proof-of-principle f f i i l d decisions i i Performance Engineering-scale demonstration and optimization to desired levels of performance Demonstration MidMid to large-scale large scale system demonstration [Commercialization] Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 16 8 VHTR - Technology gaps Process-specific R&D gaps exist to adapt the chemical process and the nuclear heat source Qualification of high-temperature high temperature alloys and coatings coatings. Producing hydrogen using the I-S process Performance issues include development of a highperformance helium turbine. Modularization of the reactor and heat utilization systems is another challenge for commercial deployment of the VHTR. Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 17 Concept of a modular HTR (1/3) Fuel element and restraining the fission products up to 1600°C Brennelement Kr - Freisetzung 10 -1 SiC Beschichtung 85 Graphit 10 0 10000 bis 20000 coated particles Graphit SiC 10 -7 1300 1500 1700 1900 2100 2300 2500 Temperatur / °C -2 10 -4 Sr 90 Cs 137 Kr 85 10 -5 10 -6 10 -7 0 100 200 300 400 500 Heizdauer bei 1600 °C / h Graphit Spring 2011 / Prof. W. Kröger 10 -5 10 -3 Freisetzungsanteil Kern 10 -3 Safety of Nuclear Power Plants 18 9 Concept of a modular HTR (2/3) Thermal power 300MW Circular core Pre stressed primary loop as an burst prove inclusion Interior concrete cell with a limitation on air amount Underground arrangement of the reactor building Use of the primary loop to steam production, gas turbines and process heat applications (with inbetween heat exchanger) 15 11 feedwater 12 5 steam 10 9 2 3 8 1 7 to the filter G 13 16 4 M 14 6 Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 19 Concept of a modular HTR (3/3) Underground arrangement of the reactor 18 20 15 0 0 19 12 11 5 10 Speisewasser 9 3 Frischdampf p 8 2 7 17 1 G M 16 Spring 2011 / Prof. W. Kröger 13 4 14 6 Safety of Nuclear Power Plants 20 10 Accident behaviour of a modular HTR (1/10) Accident assumptions internal Causes external Causes external causes (beyond licensing) • Crash of a plane (phantom) • Terrorist attacks with planes (Booing 747) • Gas cloud explosion • Sabotage • Massive water ingress into the primary system • Earthquake (b < 0,3 g) • Impacts pacts of o war a (missiles) ( ss es) • Massive air ingress into the primary system • Fire • Full loss of coolant • Full loss of the active residual heat removal • Extreme reactivity disturbances • Tornados, hurricanes and floods • Extreme earthquakes (b > 0,3 g) • Meteorites • Massive damage of reactor components Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 21 Accident behaviour of a modular HTR (2/10) Stabilitätsprinzip i i nichtschmelzbare Brennelemente und C Corestrukturen t kt Wirkung nukleare Stabilität selbsttätige Begrenzung von nuklearer Leistung und Brennstofftemperaturen thermische Stabilität selbsttätige Nachwärmeabfuhr aus dem Reaktorsystem chemische Stabilität selbsttätiger Schutz gegen Korrosion mechanische Stabilität Requirements for a non melting inherent safe modular HTR Fulfilment of the principles for all accidents due to internal and foreseeable external cause (requirements from licensing) Fulfilment of the principles for all accidents including unforeseeable occurrences such as terrorist attacks or extreme earthquakes selbsttätiger Schutz gegen mechanische Corezerstörung 11 Accident behaviour of a modular HTR (3/10) Automatic removal of the residual heat in the case of failing of all active heat removal (for example: PBMR, additionally the total failure of the first shut down system was assumed) max T Brennst. (°C) 1600 1400 1200 1000 800 max 600 TBrennst. 400 TBrennst. 200 0 Zeit (h) 0 20 40 60 80 120 100 Fuel elements can never melt Maximum fuel temperature remains below 1600°C Most fuel elements stay far below this temperature Total fission product release stays below 10-5 of the inventory Accident behaviour of a modular HTR (4/10) Biological analogue for the automatic residual heat removal and p limitation of accident temperature Tmax < 1600 °C QN Tw Tmax < 38 °C < 400 °C QN a A ( Tw - Tu) Q Tmax - To QN /H 4c Tmax QN 600 kW To Tw Tu Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants < 37 °C Tmax - To Q/H 4 i Tmax (entspricht der Nachwärme eines Cores mit Pth = 300 MW nach etwa 30 h ) To Q a A ( To - Tu) To Tu Q 60 W (entspricht der Wärmeabgabe eines Menschen unter “normalen” Bedingungen) 24 12 Accident behaviour of a modular HTR (5/10) New safety requirements for nuclear technology 105 t K (t) 4 1,1 10 t zukünftige Annahmen Flammentemperatur (°C) max. 1200 °C zukünftige Annahmen heutige Annahmen h ti heutige Annahmen 800 °C t 0 Schutz der Gebäude gegen Penetration und Zerstörung 30 min 2h t Schutz gegen langandauernde Flammeneinwirkungen mit hohen Temperaturen bei Kerosinbrand Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 25 Accident behaviour of a modular HTR (6/10) Reaktor Bauschutt (mit Isolationswirkung) T (°C) 1500 Core (maximum) 1000 Reaktordruckbehälter 500 0 Behälter zu 60% isoliert Behälter zu 99% isoliert 0 20 40 60 80 100 120 140 160 180 200 220 240 Zeit (h) Spring 2011 / Prof. W. Kröger Automatic removal of the residual heat after the destruction of the reactor t b building ildi (for (f example l caused by an extreme earthquake or a plane crash caused by terrorists) Reactor totally covered with rumble Residual es dua heat eat is s st still removed e o ed automatically, but slowed down Maximum fuel temperature stays below 1600°C Total fission product release stays below 10-5 of the inventory Safety of Nuclear Power Plants 26 13 Accident behaviour of a modular HTR (7/10) P/PO 6000 (%) Automatic limitation of nuclear power and fuel temperature when faced with extreme reactivity transient (total loss of the first shut down system: ρ = 1,2 % in a short period of time) 0.1 s 1.0 s 4000 Leistung des Reaktors 10.0 s 2000 t/s 0 0 10 20 Strongly negative temperature coefficient ends the excursion Transient T i t heat h t is i quickly i kl and d ffor the most part stored in the fuel element graphite Maximum fuel temperature stays below 1600°C TB 1300 (°C) 1200 1100 0.1 s 1000 1.0 s 10.0 s 900 Brennelementtemperatur 800 700 t/s 10 0 20 Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 27 Accident behaviour of a modular HTR (8/10) Concepts for minimising the consequences of an air ingress accident: acc de t Vluft 5000m3 Granulat 6 8 1 2 4 5 3 7 1) Internal concrete cell (sealed against outside air) 2) Reactor building 3) Draining channel 4) Sealing flaps (gravity) 5) Sealing plug 6) Granulate silo 7) Filter 8) Flue Spring 2011 / Prof. W. Kröger Draining the primary Helium through pipes over a filter Automatic limitation on the amount of air within the concrete cell (V < 5000 m3) by automatic closing mechanisms (flaps and pouring granulate) Limitation of possible graphite corrosion to < 500 kg C and therefore no radiological consequences Limitation of possible graphite corrosion to < 100 kg by simple intervention methods (protection of investment) Safety of Nuclear Power Plants 28 14 Accident behaviour of a modular HTR (9/10) Verhinderung der K Korrosion i von Brennelementen Begrenzung der Mengen Mengenrate an Luft Begrenzung g g der Menge an Luft Volumen der inneren Betonzelle ist begrenzt (<5000m³) Vorgespannter Reaktordruckbehälter mit kleinen Öffnungen Korrosionsbeständige SiC-beschichtete Brennelemente nach h erfolgter f l t D Druckk entlastung schließt eine Klappe oder ein Granulatvorrat den Kanal ab Berst-Schutz für die Behälter innere i B Betonzelle t ll ist mit Inergas gefüllt Intervention in innerer Betonzelle innere Betonzelle wird mit Inertgas gefüllt Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 29 Accident behaviour of a modular HTR (10/10) Comparison of the fuel behaviour after loss of the active residual heat removal. PWR 5 ~ HTR Brennstofftemperatur TB (°C) Tmelt = 2850 °C 10 Ci Brennstab ~< ZrCanning 0 S TB Brennstofftemperatur <1600 °C UO2-Kern TC ~ 600°C UO2Tablette 1Ci Partikel C 1h SiC t Schäden an den Brennelementen 30 h t C C Matrix S Schäden an den Brennelementen 1 -5 10 1h Spring 2011 / Prof. W. Kröger t 30 h Safety of Nuclear Power Plants t 30 15 Total Assessment of Safety (1/2) Integral proof of safety behaviour elektrische l kt i h Widerstandsheizung Spring 2011 / Prof. W. Kröger Simulation of the residual heat in a graphite-sphere-pile by electrical resistance i t h heating. ti ((~ 2 MW ffor a 300 MWth - Core) Analysis of all assumable accidents (total loss of coolant, air ingress, water ingress, pressure discharge, mechanical impacts on the elements of the shut down system, component damage when pressure relief) f) Validation of all calculating programs for extreme accidents (heat transfer throughout all structures, flow phenomena, acts of pressure discharge). Safety of Nuclear Power Plants 31 Total Assessment of Safety (2/2) Fuel elements can never melt; there is no heating up of the fuel elements to a temperature above 1600°C The principal of automatic residual heat removal can never fail, even after the destruction of the reactor building it stays intact. The reactor would even stay resistant against extreme reactivity transient. There is no temperature raise of the fuel elements to above 1600°C The pre-stressed reactor containment can not burst; It is impossible that an unacceptable amount of air enters the primary loop, graphite corrosion is minimal. The ingress of larger amounts of water into the primary loop will not lead to unacceptable states of the reactor; in the case of gas turbines operating this accident is not applicable. Against extreme, in future even more severe external impacts, the underground way of building with covering mound, and fast removal of spherical fuel elements as protection There is no accident in which case a significant amount of radiation is released Spring 2011 / Prof. W. Kröger Safety of Nuclear Power Plants 32 16