ML20203G254
| ML20203G254 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/23/1986 |
| From: | Bernero R Office of Nuclear Reactor Regulation |
| To: | MONROE COUNTY, MI |
| References | |
| NUDOCS 8608010043 | |
| Download: ML20203G254 (9) | |
Text
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July 23, 1986 DISTRIBltTION:
1DocketNo.350-341, NRC PDR Docket 50-341 Local POR BWD-3 r/f EAdensam DLynch County of Monroe, Michigan RBernero Board of Commissioners EHylton 106 East First Street Attorney, 0GC Monroe, Michigan 48161 JPartlow EJordan
Dear Commissioners:
BGrimes ACRS (10)
The NRC has been advised by Representative John Dingell that you are interested in the safety of the Mark I containment as is used at the Fermi-2 Plant.
In response to Representative Dingell, we have prepared a paper describing the Mark I containment ar.d the ongoing efforts to address concerns regarding that containment's ability to miticate severe accidents.
We are providing you with a copy of this paper (Enclosure 1) for your use.
In addition, I intend to discuss the containment issues at the public meeting on the Fermi site on July 28, 1986.
If you wish any further information please call me (301 492-7373) or Ms. Elinor Adensam of my staff (301-492-8180).
Sincerely, Orinigal Signed by Elinor G. Adensam for i
Robert M. Bernero, Director Division of BWR licensing Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page h
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Ferni-2 Facility cc:
Mr. Harry H. Voint, Esq.
Ronald C. Callen LeBoeuf, Lamb, Leiby A MacRae Adv. Plannino Review Section 1333 New Panpshire' Avenue, N. W.
Michigan Dublic Service Commission Washington, D. C. 70036 6545 Mercantile Way P. O. Box 30??1 John Flynn, Eso, lansing, Michiaan 48909 Senior Attorney The Detroit Edison Company Pegional Administrator, Peoion TII 2000 Second Avenue U. S. Nuclear Pecultitory Commission Detroit, Michigan 48226 790 Porsevelt Road Glen Ellyn, Illinois 60137 Mr. Dennis R. Fahn, Chief Nuclear Facilities and Environmental Monitorina Section Office Division cf Padioloaical Health P. O. Fox 30035 Lansinc. Michican 48909 Pr. Robert Voolley Actina Supervisor-Licensing The Detroit Edison Corpany Fenri Unit 2 64n0 No. Dixie'Highwev Newoort, Michican 48166 Fr. Walt Rogers U. S. Nuclear Regulatory Connission Resident Insnector's Office 6450 W. Dixie Highway Newport, Michicar t8166 Montce County Office of Civil Preparedness 963 South Daisinville Monroe, Michigan 48161 Mr. B. Ralph Sylvia Group Vice President Detroit Edison Company 6400 North Dixie Highway Newport, Michigan 48166 36
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SEVERE ACCIDENT SAFETY IN BOILING WATER REACTORS WITH MARK I CONTAINMENT As the name indicates, a boiling water reactor (BWR) is a reactor in which the I
t'ater fed to the reactbr core boils right there in the reactor vessel and then passes as steam directly out to the turbine generator where its energy is converted to electricity.
The exhausted steam, after condensation, is I
returned to the reactor as feedwater.
Figure 1 shows a simple schematic of a BWR plant.
The reactor is enclosed in a special containment structure.
The feedwater enters and the steam leaves this containment structure through multiple, large diameter pipes equipped with redundant valves which can be closed in an emergency.
In the pressure suppression containment which is used in all large U.S. BWRs, a very large quantity of water, up to one million gallons, is storea in a special compartment of the containment called the suppression pool.
Many auxiliary and emergency cooling systems are provided
?.o pump cooling water into the reactor and to cool the containment atmosphere and its suppression pool.
If a pipe breaks by accident, the containment closes to isolate the reactor in the containment and many cooling systems are called into play to cool the reactor and the suppression pool, removing the stored energy and heat generated by radioactive decay.
Thus, the BWR is an open system removing large quantities of energy to nearby l
Gquipment which, in emergencies, converts to a closed system, basically relying on external cooling of the containment to remove the bottled-up energy.
The most common type of pressure suppression containment in the U.S.
is the Mark I type shown in Figure 2, which is used in the 24 U.S. BWRs listed
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in Table 1.
The reactor is contained in the drywell portion of the containment, shaped like an electric light bulb standing upside down.
The suppression pool partially fills a toroidal shell around the base of the
" bulb" and a series of ducts is installed to guide steam and other releases into the suppression pool which quenches the steam and also absorbs much of the radioactive material (except gases).
" Severe accidents" is the term most commonly used to describe accidents in whichtheheactorcoreisseverelydamaged.
As happened at Three Mile Island, prolonged loss of core cooling can allow the heat of radioactive decay in the
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TABLE 1 BOILING WATER REACTORS WITH MARK I CONTAINMENTS LICENSED OPERATING
- PLANTr, POWER LICENSE NAME LEVEL DATE COUNTY STATE UTILITY 6
BROWNS FERRY 1 3293 12/20/73 LIMESTONE COUNTY AL TVA BRGWNS FERRY 2 3293 08/02/74 LIMESTONE COUNTY AL TVA
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BROWNS FERRY 3 3293 08/18/76 LIMESTONE COUNTY AL TVA BRUNSWICK 1 2436 11/12/76 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT BRUNSWICK 2 2436 12/27/74 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT COOPER 2381 01/18/74 NEMEHA COUNTY NE NEBRASKA PUBLIC POWER DISTRICT DRESDEN 2 2527 12/22/69 GRUNDY COUNTY IL COMMONWEALTH EDISON DRESDEN 3 2527 03/02/71 GRUNDY COUNTY IL COMMONWEALTH EDISON DUANE ARNOLD IGb8 02/22/74 LINN COUNTY IA IOWA ELECTRIC POWER & LICHT FERMI 2 3292 07/15/85 MONROE COUNTY MI DETROIT EDISON FITZPATRICK 2436 10/17/74 OSWEGO COUNTY NY POWER AUTHORITY OF STATE OF NY HATCH 1 2436 10/13/74 APPLING COUNTY GA GEORGIA POWER HATCH 2 2436 06/13/78 APPLING COUNTY GA GEORGIA POWER HOPE CREEK 1 3293 04/11/86 SALEM COUNTY NJ PUBLIC SERVICE ELECTRIC & GAS MILLSTONE 1 2011 10/16/70 NEW LONDON CT NORTHEAST NUCLEAR ENERGY MONTICELLO 1670 01/19/71 WRIGHT COUNTY MN NORTHERN STATES POWER NINE MILE POINT 1 1850 08/22/69 OSWEGO COUNTY NY NIAGARA MOHAWK POWER OYSTER CREEK 1 1930 08/01/69 OCEAN COUNTY NJ GPU NUCLEAR CORP PEACH BOTTOM 2 3293 12/14/73 YORK COUNTY PA PHILADELPHIA ELECTRIC PEACH BOTTOM 3 3293 07/02/74 YORK COUNTY PA PHILADELPHIA ELECTRIC PILGRIM 1998 06/08/72 PLYMOUTH COUNTY MA BOSTON EDISON QUAD CITIES 1 2511 12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON QUAD CITIES 2 2511 12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDISON VERMONT YANKEE 1593 02/02/73 WINDHAM COUNTY VT VERMONT YANKEE NUCLEAR POWER i
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core to build up to the point that the fuel begins to disintegrate, the zirconium metal cladding melts or reacts with residual steam to form combustible hydrogen,'and even the ceramic uranium oxide fuel pellets can
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melt.
A great deal of attention is being given to understanding the behavior of reactors and their containments in severe accidents, especially since the Three Mile Island accident.
The objectives are to ensure that the likelihood of core melt accidents is very low and that, should one occur, there is substantial assurance that the containment will mitigate its consequences.
The severe accident behavior of a BWR with a Mark I containment, the Peach Bottom plant, was assessed in the Reactor Safety Study (WASH-1400 or NUREG-75/014) which was published in 1975.
That study indicated a relatively low overall risk for the BWR, principally due to its ability to prevent core melt.
The containment was estimated to provide very little mitigation of core melt consequences because the buildup of pressure under accident conditions would be a direct cause of containment failure unless adequate cooling was preserved.
Consistent with operating procedures in place in 1975, the Study assumed little effort by the reactor operators which might effectively prsserve the containment's integrity.
The situation, more than ten years later, is different and still changing for tha better.
It is recognized today that molten core materi.al melting into the ground through the thick containment base is not the principal threat; rather, it is an atmospheric release of radioactive material which is the principal threat.
The principal factors which can cause containment failure with atmospheric release are hydrogen ignition, gas overpressure buildup to rupture, and direct attack of the drywell by core melt debris. The general situation for each of these is summarized as follows:
Hydrogen Ignition Recognizing that combustible hydrogen can be generated and released in severe cccidents, all Mark I containments now are purged and filled with inert nitrogen gas during operation so that even if hydrogen gas is formed it has insufficient cxygen available to support combustion.
Remaining questions in this area relate to how long the containment may be without this inert atmosphere in order to
. permit inspections, and how air night leak in or hydrogen leak out to nearby rooms under accident conditions Overpressure Failure Careful analysis indicates that a typical Mark I containment can withstand pressures of more than twice the design pressure without rupture.
Nevertheless, severe accidents in the extreme can generate such pressures and cause containment rupture.
Overpressure damage control procedures have been developed for pressure suppression containments and are already in place for operator use.
With these procedures the containment remains closed for most accident conditions; but, if overpressure failure threatens, large vent valves above the suppression pool chamber are opened so that the excess pressure is released gradually by bubbling the releases through the pool, forming a filtered vent containment system.
With this path assured, virtually nothing but the noble gases are released.
The radioactive no'ble gases pose a modest onposure threat offsite only in the area very close to the plant.
A number of questions are being pursued in this area. All the plants have suitably large vent valves and ducts but they vary one to another in the ability to open these valves under accident conditions.
The valves are designed for highly reliable closure, not opening.
Consideration is being given to modifying valve controls.
In addition, the vent ductwork downstream of the valves may warrant modification.
In most plants it is fairly light gauge ductwork and might be breached in accident venting.
If so, consideratio'n is being given to the effects of secondary release of radioactive gas, hydrogen, and perhaps steam into the reactor building.
Direct Attack The core melt debris, since it has melted through the reactor vessel into the drywell may, by direct radiation of heat, cause failure of connections in the drywell shell; or the debris, if sufficiently fluid, may flow out to the wall and melt through the steel. The Mark I containments have one or more spray systems in the drywell which are able to spray water along the walls and onto the floor of the drywell inhibiting direct attack.
Concerns in this area are in three general areas:
core debris modeling, shell and concrete attack modeling, and spray reliability.
In the first area, it is recognized that a molten reactor core, to melt through the bottom of a BWR, must dissolve a very i
1
.. large amount of inert metal in the lower reactor vessel, probably diluting the core melt.
The key question is whether the melt would come out moving sluggishly like Hawaii'an volcano lava or as a hot free flowing liquid.
The latter is the more threatening condition.
If core melt debris reaches the concrete floor and steel shell of the wall, it is important to understand that the path to the outside that might be opened bypasses the beneficial scrubbing of radioactive material passing through the pool.
As noted earlier all these plants have drywell spray systems, but they are designed as a secondary mode of operation for a reactor safety system.
Strong consideration is being given to enabling hookup of these systems to fire protection systems so that spray capability is almost always available.
Substantially different emergency operating procedures and training were put in place at all reactors after the Three Mile Island sccident; further improvements in these procedures are still being made.
For the Mark I containments both industry and NRC studies are being used to identify the best combined strategy for procedures and perhaps some changes in equipment such as alternate vent paths, or improved valve operability.
The Mark I studies are being given highest priority by the NRC staff and the industry. The expectation is that, with modest improvements of this type, one can achieve substantial assurance of core melt consequences mitigation by a Mark I containment.
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