THE FUKUSHIMA ACCIDENT: CAUSES, RESULTS AND IMPLICATIONS FOR EUROPE

Transcription

1 THE FUKUSHIMA ACCIDENT: CAUSES, RESULTS AND IMPLICATIONS FOR EUROPE Steven C. Sholly, Senior Scientist Institute of Safety and Risk Sciences Department of Water, Atmosphere & Environment University of Natural Resources & Life Sciences Course (Winter 2011) 24 Oct-11 Lecture 2: Fukushima Accident Progression 1

2 Introduction 2

3 Purpose of This Lecture This lecture will address two main topics. First, we will discuss the progression of the accidents that took place at Fukushima Daiichi Units 1-4. Second, we will discuss the similarities (and differences) between the Fukushima Daiichi Units and European nuclear power plants (NPPs). Before proceeding to these topics, however, we will discuss the course requirements, which discussion we deferred in the first lecture due to time constraints on lecture room availability. 3

4 Course Requirements 4

5 Course Requirements 1 The Lecture Course is open to all students at Boku and the University of Vienna. Lecture attendance is a requirement. Attendance at lectures will account for 50% of your grade. There will be Lecture Sessions 10 October, 24 October, 7 November, 21 November (needing an alternate, either 14 November or 28 November perhaps), 5 December, and 2 January. There will be Sessions on 16 January and 30 January at which presentations by lecture participants will be made and discussed (and at which attendance is also a requirement). 5

6 Course Requirements 2 Electronic copies of lecture viewgraphs will be available to class participants either at the lecture or on the following Monday by . Obtaining the viewgraphs is not a substitute for attending the lecture. Lecture participants are responsible for material covered in the viewgraphs as well as that covered orally and visually in the lectures. If you miss a lecture: (1) contact me for the viewgraphs; and (2) discuss the lecture with me and/or one or more classmates so you know what was covered orally and visually in the lectures. 6

7 Course Requirements 3 All lecture participants will make a 15- minute presentation, the quality of which represents 30% of your grade. A copy of your presentation must be submitted electronically to me not later than the day of the presentation. The topics of the presentations will be related to some aspect of the Fukushima accidents, to be proposed by the class participants and agreed to with me not later 5 December. 7

8 Course Requirements 4 There will be two examinations ("quizzes") covering materials presented in the lectures. The quizzes will account for 20% of your grade. The quizzes will take place in the Weeks of 7 November and 5 December. The quizzes will consist of multiple choice questions. The purpose of the quizzes is to ensure that you have understood the basic points presented. 8

9 Fukushima Accident Progression 9

10 The Fukushima Reactors Fukushima Daiichi is a 3.5 km 2 nuclear power station consisting of six units General Electric (GE) boiling water reactors (BWRs). Unit 1 is a 460 MWe BWR/3 Mark I design. Units 2-5 are 784 MWe BWR/4 Mark I units. Unit 6 is an 1100 MWe BWR/5 Mark II unit. We will focus on Units 1-4 since reactor damage (Units 1-3) and containment/reactor building damage (Units 1-4) occurred at these units. The next slide shows a line drawing of a Mark I pressure suppression containment. 10

11 BWR Mark I Containment 11

12 BWR/3-4 Safety Systems 1 The principal safety systems of BWR/3 and BWR/4 reactors are the following: The steam-driven Reactor Core Isolation Cooling (RCIC) for emergency cooling at high pressure (except Unit 1 has an isolation condenser or IC for this purposes instead of RCIC). The redundant steam-driven High Pressure Coolant Injection (HPCI) system for the same purpose. The motor-driven Low Pressure Core Spray (LPCS) system for emergency cooling at low pressure. The motor-driven Low Pressure Coolant Injection (LPCI) system for the same purpose. The motor-driven Residual Heat Removal (RHR) system, which removes heat from the reactor or the suppression pool and transfers it to the ultimate heat sink. 12

13 BWR/3-4 Safety Systems 2 The principal safety systems of these BWRs are (continued): RHR is an operating mode of the LPCI system. LPCI can also be used to spray water into the drywell atmosphere and into the air space above the suppression pool. The suppression pool is used to condense steam from the drywell and from the main steam lines, and also serves as a source of water for the RCIC, HPCI, LPCS, and LPCI when the condensate storage tank inventory is depleted. The next slide shows two General Arrangement drawings of Fukushima Unit 1. 13

14 Fukushima Unit 1 General Arrangement Drawing Note the Isolation Condensers on Section A-A at 31 meters above grade. Note also the location of the Spent Fuel Pool on Section B-B beginning at 23.1 meters above grade. 14

15 BWR/3-4 Safety Systems 3 The principal safety systems of these BWRs are (continued): Emergency diesel generators (EDGs) to supply electric power to plant systems when offsite power is not available. There are 10 water-cooled EDGs and 3 air-cooled EDGs at Fukushima Daiichi. Units 1, 3 & 5 have 2 water-cooled EDGs each. Units 2 & 4 each have 1 water-cooled EDG and 1 air-cooled EDG. Unit 6 has 2 water-cooled EDGs and 1 air-cooled EDG. The water-cooled EDGs are in the turbine hall basements; the air-cooled EDGs are behind the reactor building (Units 2 & 4) or in the reactor building (Unit 6). 15

16 BWR/3-4 Safety Systems 4 The principal safety systems of these BWRs are (continued): The Seawater Cooling System cools the water-cooled EDGs, the RHR heat exchangers, and other important systems. The primary containments at Units 1-5 consist of a pear-shaped (or "hanging lightbulb") drywell (containing the reactor pressure vessel) and a donutshaped torus (containing the suppression pool or wetwell). The torus is connected to the drywell by a series of vent pipes, which connect to a common header from which downcomers extend below the water surface. The vent pipes are contained in expandable bellows to allow for thermal expansion & contraction. 16

17 Mark I Drywell The drywell is a steel shell (30 mm thick) backed in most areas by nearly two meters of reinforced concrete. The exceptions are the closure head at the top, the drywell vents leading to the torus, personnel and equipment hatches, and piping & electrical penetrations. There is one double door personnel airlock and two larger bolted equipment hatches provided for drywell access. The drywell closure head is constructed with a "double tongue and groove seal". Bolts secure the drywell head to the cylindrical section of the drywell. Shielding over the top of the closure head is provided by removable reinforced concrete shield plugs. 17

18 Mark I Wetwell The wetwell (torus) is not backed by reinforced concrete. It is strictly a steel structure, supported from below. Access to the suppression chamber is provided by two manways with double gasketed bolted covers. 18

19 Mark I Reactor Building The Mark I reactor building (often referred to as a secondary containment, but this is technically incorrect) surrounds the primary containment with a lightly reinforced concrete structure below the refueling deck, and a standard industrial building at the refueling deck and above. The area above the refueling deck is equipped with automatically opening louvers to equalize pressure (a tornado design feature). Both the compartments surrounding the primary containment and the refueling deck area are served by the Standby Gas Treatment System (SGTS) which under accident consitions filters and releases air from the secondary containment to the environment via the 120 m high exhaust stack. 19

20 BWR/3-4 Main Steam Lines There are four Main Steam Lines (MSLs) which have Main Steam Isolation Valves (MSIVs) inside & outside containment. Inside containment, the MSLs are equipped with Safety/Relief Valves (SRVs) to relieve pressure to the suppression pool when the MSLs are isolated by MSIV closure. In case of a severe accident, in the main steam lines the only thing separating the reactor core from the environment are the MSIVs. The MSIVs are not perfect, and do leak under the best of conditions. This leads to radioactivity releases in all BWR severe accidents. 20

21 Main Steam Isolation Valves 21

22 Inerted Primary Containment The primary containment (both wetwell and drywell) are inerted with nitrogen (to an oxygen concentration of less than 4%) to prevent hydrogen combustion within the primary containment. It should be noted that in the U.S. the NRC allows BWRs to begin de-inerting up to 24 hours before a planned outage, and allows BWRs a 24-hour grace period to complete inerting during restart. This is not permitted in Japanese BWRs 22

23 Inerted Primary Containment 23

24 The Fukushima Accidents 24

25 Accident Progression 1 You will recall from the first Block Lecture that at the time of the earthquake, Fukushima Daiichi Units 1-3 were at full power. Unit 4 was shut down and defueled (with the full reactor core offloaded to the spent fuel pool). Units 5 & 6 were shut down and refueled, awaiting restart. When the earthquake struck, Units 1-3 scrammed (shut down) automatically, and offsite power from six separate sources was lost (4 lines at 275 kv and 2 lines at 500 kv). The earthquake also damaged at transformer station 10 km from the site, so that even though the grid system was restored in 50 minutes, offsite power was still not able to be connected to the plant site. 25

26 Offsite Power Line Failure Example 26

27 Accident Progression 2 All of the 13 available EDGs (one additional EDG was out of service for maintenance) automatically started and provided emergency power. Plant safety systems functioned normally. When the tsunami waves struck the site 41 minutes after the earthquake, the waves reached more than 14 meters above MSL (more than 6 meters above design). The inundation level was more than 4 meters above the grade level at Units 1-4, and more than 2 meters above the grade level at Units 5 & 6. The Fukushima Daiichi site was flooded (including all six units, a common spent fuel storage pool, and a dry spent fuel storage facility). 27

28 Accident Progression 3 Of the 13 available EDGs at the site (one was out of service for maintenance), 9 were located in sub-basement levels of the turbine halls and were immediately flooded by the tsunami (2 air-cooled diesels were at grade level behind Unit 4, and 2 were in the Unit 6 reactor building). Even if the diesels themselves had not been flooded, the sea water pumps that provided cooling to the water-cooled EDGs were destroyed by the tsunami waves. In addition, the diesel fuel tanks for Units 1-4 were destroyed by tsunami waves. Units 1-4 were in station blackout (total loss of AC power), with only battery (DC) power available. There was no cooling available to the spent fuel pools. Units 5 & 6 shared a single air-cooled EDG. 28

29 Fukushima Site Before Tsunami Impact 29

30 Fukushima Site With Tsunami Wave Approaching (Units 5-6 Diesel Tanks on Left) 30

31 Fukushima Tsunami Impact 1 31

32 Fukushima Tsunami Impact 2 32

33 Fukushima Tsunami Inundation Area 33

34 Fukushima Site After Tsunami Impact 34

35 Damage Details, Unit 3 Seawater Pumps 35

36 Damage Details, Unit 5 Seawater Pumps 36

37 Accident Progression 4 Once coolant injection fails (e.g., stopping of RCIC pump due to battery depletion), you have a situation similar to that depicted in the next slide. The core is still fully covered by water. In the next step, without further injection, and without containment heat removal, the water inventory in the reactor vessel begins to boil off. The steam produced is directed to the suppression pool via the safety/relief valves (SRVs). 37

38 Status While RCIC Still Operates 38

39 RPV Inventory Lost by Boiloff Thru SRVs 39

40 Accident Progression 5 Eventually, unless injection is restored, the top of the fuel assemblies in the reactor core start to uncover. The fuel begins to heat up as the reactor vessel water level drops lower and lower. Finally, the fuel cladding starts to fail, and the socalled "gap release" occurs (accounting for about 5% of the noble gases, iodine, and cesium in the fuel rods). With 75% of the fuel exposed, an exothermic metal/water reaction begins, and accelerates core uncovery. 40

41 Top of Fuel Assemblies Exposed 41

42 Half of Fuel Assemblie Length Exposed 42

43 75% Core Uncovery, Zr/H20 Reaction Produces Heat & Hydrogen 43

44 Fuel Rod Details (Intact, L; Failed, R) 44

45 Accident Progression 6 Core debris begins to form, with a molten mass forming in the core. Containment pressurization due to boiling water in the suppression pool results in a need to vent the wetwell airspace. In the extreme, venting of the drywell can be done. Venting releases radioactivity and hydrogen to the upper reactor building. If venting is not done or is not successful, the drywell heat flange will leak. 45

46 Molten Core Debris Forms 46

47 Containment Pressurization & Venting 47

48 Hydrogen Burn/Detonation, Reactor Building Destruction 48

49 Accident Progression 7 Eventually sufficient core debris relocates to the lower RPV head, raising the potential for RPV failure to occur. Failure of the RPV can be a local failure at a bottom head penetration, or melt-through can occur along the side-wall of the vessel. If RPV failure occurs at high pressure, the liquid core debris is ejected from the vessel ("High Pressure Melt Ejection"). 49

50 Lower RPV Head Details 50

51 Accident Progression 8 Ejected core debris rapidly gives up its heat to the drywell atmosphere, suddenly increasing both pressure and temperature ("Direct Containment Heating", DCH). If the containment has already failed at this point, mixing of air with the finely divided core debris can result in a large hydrogen burn or detonation. If the containment is intact, the rapid increase in pressure and temperature can result in drywell closure head failure, or failure of the torus in the wetwell airspace. 51

52 Accident Progression 9 Once the RPV pressure is relieved through the failure location, further core debris melts over a period of hours and drops into the drywell pedestal area under the vessel. This core debris heats and chemically attacks the concrete on the drywell floor. During HPME, ejected core debris can contact the drywell shell. If it does so in the absence of water on the drywell floor, melt-through of the drywell shell is likely, providing another path for radioactivity release. 52

53 Drywell Pedestal Area (Peach Bottom) 53

54 Drywell Vent Opening (Browns Ferry) 54

55 Interior View of Suppression Pool 55

56 Accident Progression 10 If there is no water to cool the core debris on the floor of the drywell, it is possible for core debris (depending on the height of the vent pipes and the quantity of core debris involved) to flow into the vent pipes, melt through them, and deposit on the lowest level of the reactor building. Then so-called "Molten Corium/Concrete Interaction" (MCCI) can take place in both the drywell and the reactor building. 56

57 Severe Accident Mitigation Measures at Fukushima 57

58 Unit 1 Hydrogen Explosion - Deflagration 58

59 Unit 1 Before & After Hydrogen Explosion 59

60 Unit 3 Hydrogen Explosion - Detonation 60

61 Unit 3 Before & After Hydrogen Explosion 61

62 Unit 3 Dark "Smoke" from Molten Corium/Concrete Interactions (MCCI) 62

63 Unit 4 Possible Hydrogen Path 63

64 Unit 4 Before & After Hydrogen Explosion 64

65 Questions About Unit 4 (1) Japanese government reports suggest that hydrogen was transported from Unit 3 through the stack release ductwork, and flowed backwards into Unit 4. But why didn't Unit 4 explode when Unit 3 detonated? And note the ductwork at Unit 3 was blown apart in the Unit 3 hydrogen detonation. Suggestion from NGOs and independent sources is that the Unit 4 spent fuel pool was the source of the hydrogen BUT... 65

66 Questions About Unit 4 (2) Instead of finding the Unit 4 spent fuel pool dry (as suggested by the Chairman of the NRC), the spent fuel pool was found to have the fuel covered by water that was boiling/steaming off. There is no visible evidence of fuel damage (i.e., no warping of fuel channels, no warping of fuel assemblies, no charring of pool walls due to hydrogen combustion, no settling of fuel assemblies, pool water samples provide no evidence of fuel damage, etc.). See next slide. And note the Unit 4 hydrogen explosion (6 am, 15 March) was 19 hours after the Unit 3 detonation (11 am, 14 March). 66

67 Unit 4 Spent Fuel Pool After Explosion 67

68 Status of Units 1-4 After the Accident 68

69 Similarities of Fukushima Units With European NPPs 69

70 Similar Units in Europe There are certain similarities between the Fukushima BWR/3 and BWR/4 Mark I units and nuclear power plants in Europe. The similar reactors include General Electric BWRs, Asea Atom BWRs, Siemens BWR/69 & BWR/72 reactors, and VVER-440/213 PWRs. These similarities will be discussed on the following viewgraphs. 70

71 GE BWRs in Europe 1 There are a small number of General Electric BWRs operating in Europe: Santa María de Garoña and Cofrentes in Spain. Mühleberg and Leibstadt in Switzerland. Santa María de Garoña is a BWR/3 Mark I facility similar to Fukushima Daiichi Unit 1. This unit is scheduled to close in Mühleberg is a BWR/4 Mark I unit, with some similarities to Fukushima Units 2-4, but with major design differences. This unit is scheduled to close in

72 GE BWRs in Europe 2 Mühleberg is unique among BWR/4 units in the world: Mühleberg has a full double pressure suppression containment (steel-lined, reinforced concrete), with an outer suppression pool containing a filtered venting device (multiventuri scrubber system, MVSS) and a connection of the outer suppression pool airspace to the 125-meter high elevated stack. Mühleberg has a bunkered safety system complex called SUSAN. The SUSAN building houses 2 of the 3 EDGs, a redundant RCIC train, a 2-train redundant Alternate Low Pressure System or ALPS, a 2-train redundant torus cooling system. 72

73 GE BWRs in Europe 3 Mühleberg also has: The SUSAN complex is a bunkered, flood-proofed, reinforced concrete structure with its own control room. Emergency power is also available from the nearby Wasserkraftwerk Mühleberg. An elevated water tank ("Hochreservoir") on a nearby hill is available to drain water by gravity for emergency coolant injection or drywell spray purposes. The containment free volume, wetwell water contents, and the condensate storage tank are twice as large (considering the plant power level) as they are in the typically referenced Peach Bottom plant in the U.S. 73

74 GE BWRs in Europe 4 Leibstadt and Cofrentes are BWR/6 Mark III units with a different containment concept than the Mark I (see the next viewgraph). Cofrentes' license expires in Leibstadt is unique among BWR/6 Mark III units in having a full double pressure suppression containment, as well as a bunkered safety complex and a filtered venting system. Leibstadt is scheduled to close in

75 GE Mark III Containment 75

76 ASEA-ATOM BWRs ASEA-ATOM produced BWRs that are operating in Finland (Olkiluoto 1 & 2) and Sweden (Forsmark 1, 2 & 3; Oskarshamn 1, 2 & 3; Ringhals 1). See next viewgraph. Siemens produced two lines of BWRs (BWR/69 and BWR/72) in Germany. All of the BWR/69 units are now closed. Only the BWR/72 units at Gundremmingen B & C remain in operation. 76

77 ASEA-ATOM BWR Containment 77

78 VVER-440/213 (1) Now wait a minute, you might say how could a Soviet-designed PWR be similar to General Electric BWR? There are two main similarities. First, the VVER-440/213 uses a pressure suppression system. Second, the reactor building (confinement) has similarities to the BWR Mark I and Mark II drywells. See the next viewgraph. 78

79 VVER-440/213 Design 79

80 VVER-440/213 (2) The VVER-440/213 has a reactor shaft that is connected to the refueling floor, and capped by closure cap similar in concept to the BWR Mark I and Mark II drywell closure head. In addition, the pressurizer extrudes above the refueling deck in its own compartment. VVER-440/213 units are operating in the Czech Republic (Dukovany 1-4), the Russian Federation (Kola 3 & 4), Slovakia (Bohunice V2 3 & 4, Mochovce 1 & 2), and Ukraine (Rovno 1 & 2). Two more units are under construction at Mochovce. 80

81 Next Lecture 81

82 Next Lecture Block Lecture 3 will take place on 07 November from 18:30 20:00. We will discuss the Japanese Government's two official reports on the Fukushima accidents, as well as the IAEA's "Fact- Finding" Mission report. Fukushima_ pdf cn200_final-fukushima-mission_report.pdf 82

83 Contact Information 83

84 Contact Information Direct line: Secretariat: Institute URL (Deutsch): Course URL (English): erson_nr=&sprache=2 Course URL (Deutsch): Lecture Details (date, room, time): 0&clvnr=