Nuclear Safety. Lecture 3. Beyond Design Basis Accidents Severe Accidents

Transcription

1 Nuclear safety Lecture 3. Beyond Design Basis Accidents Severe Accidents Ildikó Boros Prof. Dr. Attila Aszódi Budapest University of Technology and Economics Institute of Nuclear Techniques (BME NTI) 1

2 Levels of defence in depth Mitigation of radiological consequences of significant off-site releases of radioactive materials Control of severe plant conditions including prevention of accident progression and mitigation of severe accident consequences Control of accident within the design basis Control of abnormal operation and detection of failures Prevention of abnormal operation and failures Conservative design and high quality in construction and operation Control, limiting and protective systems and other surveillance features Engineered safety features and accident procedures Complementary measures and accident management Off-site emergency response Nucear safety Source: IAEA 2 09/05/2017

3 Beyond Design Basis Accidents (BDBA) Postulated Initiating Event (PIE, leading to DBA) + coincidental failure of redundant safety systems (e.g.: no available ECCS during a SB LOCA) or extreme rare events At different reactor designs, different extent of accidents Examples: ATWS Anticipated transient without scram (In normal mode, during scram the safety rods are falling by gravity into the core after the EM release. If EM fails to release, the emergency shutdown is not successful. Or: control rod stuck) Total loss of feedwater (and emergency feedwater) system Total loss of ultimate heat sink Total loss of electricity (including EDGs) LUHS SBO 3

4 Beyond Design Basis Accidents (BDBA) Possible operator actions for mitigation Primary feed and bleed: blowdown through pressurizer relief valve, injection from ECCS Secondary feed and bleed: blowdown and (emergency) feedwater injection at secondary side (also in case of DBA events) Containment isolation Containment depressurization (via containment spray system or filtered containment venting or outer cooling of containment) 4

5 Severe accidents The consequence is core degradation / melting and / or significant radioactive release into the environment Basic processes: In-vessel processes 1. Core degradation 2. Steam explosion 3.Induced breaks Vessel failure 4. Losing vessel integrity Ex-vessel processes 5. Direct heating 6. Hydrogen explosion 7. Corium-concrete interactions Source: IAEA 5

6 In-vessel phase core degradation Core heated by decay heat after shutdown Typical temperatures 1130 o C metallic uranium meltpoint, formation of melted U during UO 2 /Zr reaction 1200 o C acceleration of Zr oxidation 1300 o C formation of Zr/steel eutectic 1450 o C steel meltpoint 1845 o C Zr meltpoint 1970 o C α-zr(o) meltpoint 2600 o C (U,Zr,O) ceramic melt 2690 o C ZrO 2 meltpoint 2850 o C UO 2 meltpoint Remanent heat power (%) Time after shutdown (min) Remanent heat power in VVER-440 (MW) 6

7 Core degradation oxidation Zr-steam reaction above 1200 o C Zr + 2 H 2 O ZrO 2 + 2H 2 + Q Q= 5 MJ/kg Zr From 1 kg Zr about 0.5 m 3 H 2 (STP) can be produced (it means about 960 kg H 2 in case of a 900 MW PWR Highly exoterm reaction, reaction velocity depends on temperature Heat-up of the core accelerates, reaching the Zr meltpoint Source: Dr. Hózer Zoltán, KFKI AEKI 7

8 Core degradation CODEX experiment CODEX, EK: Severe accident conditions, high temperature 600 mm long fuel assemblies, heated by electrical heating Adjustable coolant parameters Source: Dr. Hózer Zoltán, KFKI AEKI 8

9 Core degradation CODEX CODEX, EK: experiment Source: Dr. Hózer Zoltán, KFKI AEKI 9

10 Core degradation PHEBUS experiment Experiments on PHEBUS research reactor, at Cadarache, France Real fuel bundle in closed channel Even spent NPP fuel can be used For the determination of release of fission products, simulation validation 10

11 Core degradation PHEBUS experiment Change of fuel temperature and core degradation in PHEBUS facility Source: Florian Fichot, IRSN 11

12 Core degradation PHEBUS experiment Radiography of degraded fuel bundle in PHEBUS facility (Source: Belpomo, CEA) 12

13 Core degradation corium migration After mobilization, corium appears in other fuel assemblies Crust surrounding the corium can not withstand effect of corium crust failure Corium relocated to vessel bottom in few minutes Corium can flow at outer part of core, or at the place of the already missing FAs Crust failure and corium relocation in the RPV (Source: Florian Fichot, IRSN) 13

14 Steam explosion in the vessel Corium-water interaction between the relocating corium and the remaining water in the bottom of RPV Possible consequences: Pressure increase in the vessel and in the primary circuit Further oxidation of Zr remaining in the core Corium Jets Molten corium Rupture of vessel bottom, corium ejection Corium ejection through the vessel top that could damage the containment In-vessel steam explosion can lead to RCS break Residual water Corium-water interaction Source: Florian Fichot, IRSN 14

15 Severe accidents The consequence is core degradation / melting and / or significant radioactive release into the environment Basic processes: In-vessel processes 1. Core degradation 2. Steam explosion 3.Induced breaks Vessel failure 4. Losing vessel integrity Ex-vessel processes 5. Direct heating 6. Hydrogen explosion 7. Corium-concrete interactions Source: IAEA 15

16 Vessel failure heat transfer in corium TMI experiences corium in separate layers: at lower layers mainly UO 2 and ZrO 2, above the melted metallic phase Melted U-Zr-O solidifies, crust is created Most of the heat transferred through the melted steel, that could lead to local damage (focusing effect) Experimental result: because of the natural convection of the corium in the vessel, the maximal heat flux can be observed at 70, minimal heat flux at the lowest point of the vessel 16

17 Vessel failure heat transfer in corium Experiment: EC-FOREVER (European Commission s failure of reactor vessel retention) facility 1/10 th scale (400 mm diameter) carbon steel vessel, constructed, welded, and heat-treated according to the vessel manufacture code. Investigated experiment: pouring an oxide melt into the vessel at 1200 C, heating the melt with a heater to maintain it at 1100 to 1200 C, then, pressurizing the vessel wall to 25 bars pressure with Argon gas 30% CaO + 70% B2O3 17

18 Vessel failure heat transfer in corium RASPLAV facility, Russia Examination of melted UO 2 / ZrO 2 / Zr A part of the melted metallic phase sinks to the vessel bottom Distribution of fission products depends on their chemical form (oxides in the ceramic phase, metals in the melted metallic phase) 18

19 TMI: 20 tons of corium was found in vessel bottom head No intensive corium-water reaction (it could have lead to 2 MPa overpressure in the vessel!) Vessel integrity remained! Vessel failure TMI experiences RPV of TMI-2 after the accident Source: Florian Fichot, IRSN 19

20 Rupture of vessel bottom: different mechanisms depending on the primary pressure and decay heat Low pressure: melting Vessel failure phenomena Intermediate pressure: creep fracture High pressure: ductile fracture Creep fracture has the highest probability Above 800 o C vessel alloy can breach at bar pressure Vessel failure needs at least several hours Vessel failure due focusing effect (corium release ~ 40-60%) Vessel failure due focusing effect and pressue peak (corium release: ~100%) Source: Lower Head Failure (LHF) Tests at Sandia National Laboratory (SNL) 20

21 Severe accidents The consequence is core degradation / melting and / or significant radioactive release into the environment Basic processes: In-vessel processes 1. Core degradation 2. Steam explosion 3.Induced breaks Vessel failure 4. Losing vessel integrity Ex-vessel processes 5. Direct heating 6. Hydrogen explosion 7. Corium-concrete interactions 21

22 Corium-coolant interaction is possible: In RPV bottom Under the RPV in the reactor cavity, if there is water at the vessel failure Containment pressure peak occurs Ex-vessel steam explosion Possible scenarios leading to steam explosion Source: Daniel Magallon, CEA Process of steam generation Source: Daniel Magallon, CEA 22

23 Direct containment heating At high pressure vessel failure corium and crust can be ejected from the vessel into the containment Containment temperature rises, pressure peak occurs Metallic part of melted core oxidizes Hydrogen generation in containment Can lead to containment integrity failure Calculated temperature at direct containment heating (Source: Berthoud, Valette) 23

24 Phenomenon depends on local hydrogen concentration: deflagration: at lower hydrogen concentration, it causes only pressure peak in the containment flame acceleration or transit to detonation: at higher hydrogen concentration, it can endanger containment integrity Hydrogen explosion 24

25 Hydrogen explosion Hydrogen production H 2 mainly from oxidation of metallic parts H 2 already present in containment (from air or from earlier processes) Hydrogen deflagration Hydrogen burning in jet (H 2 direct from corium) Propagation in the containment Explosion many parameters needed (proper H 2 and O 2 concentration) TMI About 350 kg H 2 produced (combustion heat: 120 MJ/kg) Deflagration of hydrogen caused the only significant load to the containment But pressure peak remained under the design value TMI containment kept its integrity Hydrogen deflagration in containment (M.M. Pilch, 1995) Hydrogen deflagration in TMI (W. Breitung, FZK) 25

26 Corium-concrete interaction MCCI Molten Core-Concrete Interaction: Melt-through of concrete basemat: concrete becomes plastic under the corium Non-condensing gases produced -> peak pressure in containment Processes depend on concrete ingredients: Si-concrete: fast erosion with low gas production Limestone concrete: slow erosion with significant gas production Erosion of 400 kg corium on different concrete; A=50cm*50cm; Q=150kW (Source: Hans Alsmeyer, IKET) 26

27 Severe accidents Effect of a severe accident: core melt and / or large radioactive release into the environment Basic processes: How to avoid? In-Vessel Progression 1. Core degradation 2. Steam explosion 3. Induced breaks in the primary circuit Vessel Failure 4. Vessel rupture Ex-Vessel Progression 5. Direct containment heating 6. Hydrogen explosion 7. Molten core - concrete interaction 27

28 How to avoid vessel failure? (VVER-440) Objective: keep the molten corium inside the reactor pressure vessel (in-vessel corium retention) Tool: External cooling of RPV Built-in feature in AP1000 Realized in Paks NPP Issues: Corium stratification and focusing Possible CHF on reactor outer surface Narrow flow path in reactor cavity for two phase flow Overpressurization of containment building CERES (Cooling Effectiveness on Reactor External Surface) test facility in EK, Budapest 28

29 Different tools for hydrogen mitigation Catalytic recombiners Igniters CO2 dilution Atmosphere mixing (eg. using containment sprays) Filtered containment venting Strong containment design for maximum possible loads (future containments) 29

30 Mitigation of molten core - concrete interaction VULCANO test facility (Cadarache, France) Max. 100 kg corium Heated by arc plasma (3200 o C) Thermocouples, infracameras Different basement materials On ceramic basement: Less gas production Larger spreading area Smaller corium height Faster cool-down and solidification 30

31 Mitigation of molten core - concrete interaction VULCANO test facility (Cadarache, France) Cooling of the corium in the VULCANO experiment Initial cool-down rate of the surface is 300 K/s, for the inner part of corium is 5 K/s 31

32 Mitigation of molten core - concrete interaction Spreading area of EPR Spreading of the corium on a dry surface after vessel failure and melting of melt plug Source: Areva 32

33 Mitigation of molten core - concrete interaction Spreading area of EPR Later flooding of the corium layer is possible with water from IRWST (In-containment Refueling Water Storage Tank) Source: Areva 33

34 Mitigation of molten core - concrete interaction Core catcher of VVER-1000 (Tianwan NPP) Smaller core catcher in VVER-1000 than in EPR (no spreading area) External cooling of core catcher with water by natural convection Relatively high corium level 1 year solidification time 34

35 Mitigation of effects of containment direct heating - Containment venting Source: Tepco A typical containment failure: slow overpressurization (see the Fukushima accident!) Preventing: containment depressurization Containment cooling Containment venting releasing contaminated, radioactive atmosphere from the containment into the environment by controlled, filtered way Became an issue only in 80 s At first mainly in the emergency operational procedures of BWRs NRC requirement in 1988: hardened venting for Mark I type containments Hardened venting: stronger, more resistant piping, valves, remote-controlled valves Containment venting can be performed also after core melting through the wetwell 35

36 References IAEA Safety Standrads Fundamental Safety Principles, Safety Fundamentals, No. SF-1 Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants, No. SSG-3 Specific Safety Guide Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants Specific Safety Guide, Series No. SSG-4 Deterministic Safety Analysis for Nuclear Power Plants, No. SSG- 2 Specific Safety Guide Defence in depth in nuclear safety, INSAG-10, 1996 Safety of Nuclear Power Plants: Design, No. NS-R-1 IAEA Education and Training Resources - Fundamentals and Basic Professional Training Courses Gianni Petrangeli:, 2006, ISBN-13: