1 Description of the Accident Garry R. Thomas
Downloaded by UNIV OF EXETER on December 30, 2017 | http://pubs.acs.org Publication Date: December 23, 1986 | doi: 10.1021/bk-1986-0293.ch001
Safety Technology Department, Nuclear Power Division, Electric Power Research Institute, Palo Alto, CA 94303
The TMI-2 accident occurred in March 1979. The accident started with a simple and fairly common steam power plant failure--loss of feedwater to the steam generators. Because of a combination of design, training, regulatory policies, mechanical failures and human error, the accident progressed to the point where i t eventually produced the worst known core damage in large nuclear power reactors. Core temperatures locally reached UO fuel liquefaction (metallic solution with Zr) and even fuel melt ( 3 8 0 0 - 5 1 0 0 ° F ) . Extensive fission product release and Zircaloy cladding oxidation and embrittlement occurred. At least the upper 1/2 of the core fractured and crumbled upon quenching. The lower central portion of the core apparently had a delayed heatup and then portions of it collapsed into the reactor vessel lower head. The lower outer portion of the core may be relatively undamaged. Outside of the core boundary, only those steel components directly above and adjacent to the core (≤1 foot) are known to have suffered significant damage (localized oxidation and melting). Other portions of the primary system outside of the reactor vessel apparently had l i t t l e chance of damage or even notable overheating. The demonstrated coolability of the severely damaged TMI-2 core, once adequate water injection began, was one of the most substantial and important results of the TMI-2 accident. 2
Early on the morning of March 28, 1979, at almost exactly 4:00 am, the 880-MWe Three Mile Island-Unit 2 (TMI-2) pressurized water reac tor (PWR), which was operating at nearly f u l l power (98% f u l l power or -2720 MWth), had a loss of feedwater event. What should have been a normal sequence of events responding to the loss of feedwater and r e s u l t i n g i n an uneventful shutdown of the plant d i d not proceed as planned—and the reactor core u l t i m a t e l y was badly damaged. This paper presents both a b r i e f h i s t o r y of the important aspects of the 0097^156/86/0293-^2$07,O0/O © 1986 American Chemical Society
Toth et al.; The Three Mile Island Accident ACS Symposium Series; American Chemical Society: Washington, DC, 1986.
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Description of the Accident
THOMAS
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accident that affected the core damage (1-3) and a summary of the actual damage as currently known. Background A PWR such as the Babcock and Wilcox designed TMI-2 plant has separate primary coolant and secondary coolant systems. The primary system of a PWR (Figure 1 for the TMI-2 system) i s operated at a s u f f i c i e n t l y high pressure—2200 psia (15.2 MPa) f o r TMI-2—to prevent reaching a bulk saturated water temperature and net steam formation during any normal operation. At f u l l power, the pressurized water flows upward through the core at a rate of -1.38 x 10° lbm/h (-1.74 x 10 kg/sec) and i s heated from 557°F (565 K) to 611°F (595 K). The water exits the core and reactor vessel and enters the tube side of the steam generators v i a large vessel outlet pipes, which at TMI-2 are c a l l e d the candy-cane hot legs because of t h e i r c h a r a c t e r i s t i c shape (Figure 1). At TMI-2, there are two steam generators with once-through flow on the primary s i d e , as shown for a single generator in Figure 1. Each steam generator i s fed by one hot leg and i s emptied by two cold legs which return water to the reactor vessel under the driving head of main coolant pumps. Steam to drive the turbine-generator unit i s formed in the secondary system in the shell side of the steam generators. A complete TMI-2 system schematic i s shown in Figure 2, again with only one of two loops (the A loop) shown. The A loop contains the pressurizer and i t s r e l i e f valve that play a major part in the TMI-2 accident.
Downloaded by UNIV OF EXETER on December 30, 2017 | http://pubs.acs.org Publication Date: December 23, 1986 | doi: 10.1021/bk-1986-0293.ch001
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The Accident With the loss of feedwater supply to the main feedwater pumps, caused by a condensate pump t r i p , the main feedwater pumps also t r i p . (See Figure 2 for locations of these and following reactor system components.) The system, s t i l l responding normally, t r i p s the turbine generator and starts the auxiliary feedwater pumps in an e f f o r t to maintain secondary side heat removal in the steam generators. This i s required for maintaining proper primary system temperatures and pressure u n t i l the reactor can be scramed (nuclear power shutdown) and the reactor system placed in stable standby conditions. However, in spite of starting the auxiliary feedwater pumps, no feedwater supply reaches the steam generators. Block valves in 1ine with the auxiliary feedwater pumps had been closed e a r l i e r as part of the requirements of a recently-completed test operation of the system—but the valves had inadvertently remained closed after test comp l e t i o n ; hence, no feedwater can reach the steam generators. We are s t i l l in the f i r s t few seconds of the accident. Loss of a b i l i t y for the primary system to dump i t s nearly f u l l power load r e sults in an increase in water volume in the primary system because of heating. This increase i s seen d i r e c t l y as a level increase in the pressurizer located on the A loop. The pressurizer, normally about half f u l l during ful1 power operation, acts as both the primary system accumulator plus c o n t r o l l e r of system pressure. It controls pressure by maintaining the proper pressure l i m i t s in i t s steamf i l i e d upper region through a combination of e l e c t r i c a l heaters in i t s lower w a t e r - f i l l e d portion and spray cooling in i t s upper portion.
Toth et al.; The Three Mile Island Accident ACS Symposium Series; American Chemical Society: Washington, DC, 1986.
Downloaded by UNIV OF EXETER on December 30, 2017 | http://pubs.acs.org Publication Date: December 23, 1986 | doi: 10.1021/bk-1986-0293.ch001
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T H E T H R E E MILE ISLAND ACCIDENT
Figure 1. Schematic of the TMI-2 primary system (only one of two nearly i d e n t i c a l loops shown). Reproduced with permission from Ref. 1. Copyright 1980, E l e c t r i c Power Research I n s t i t u t e .
Toth et al.; The Three Mile Island Accident ACS Symposium Series; American Chemical Society: Washington, DC, 1986.
Toth et al.; The Three Mile Island Accident ACS Symposium Series; American Chemical Society: Washington, DC, 1986. • • Steam
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