ML20211H552
| ML20211H552 | |
| Person / Time | |
|---|---|
| Issue date: | 06/17/1986 |
| From: | Liaw B Office of Nuclear Reactor Regulation |
| To: | Bernero R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20209C630 | List: |
| References | |
| FOIA-87-10 NUDOCS 8606240086 | |
| Download: ML20211H552 (8) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION 7
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MEMORANDUM FOR: Robert M. Bernero, Director Division of BWR Licensing .
THRU: Gus C. Lainas, Assistant Director Division of BWR Licensing /
FROM: B. D. Liaw, Chief Engineering Branch j Division of BWR Licensing
SUBJECT:
PERFORMANCE OF BWR CONTAINMENTS IN SEVERE ACCIDENTS 4
This responds to your memorandum, dated May 29, 1986, requesting thoughtd on dealing with the performance of BWR containments in severe accidents. After -
the staff meeting when the possible " similarities" between BWR suppression pool type containments and the Chernobyl design were mentioned and shortly before receiving your May 29 memorandum, I ask C.P. Tan and Jim Lombardo of my staff to take a quick look at the potential implication in terms of over-pressure resisting capability and the purge / vent potential for preventing
" burst" mode failure. A summary of their findings is attached for your information.
Although it does not address fully all technical points made in your memorandum, the summary prcvides some relevant facts for all BWR containments including the numbers of specific design and their design pressures. Their basis conclusion supports the thrust of your proposed policy. However, it is
, not clear in our minds why Mark I's are singled out for "special treatment,"
i other than the fact it has a smaller volume. In terms of overpressure-resisting capability, Mark I's are not worse than either Mark II's or Mark III's. Additionally, we question the adequacy of the existing purge / vent valve logics for your intended purpose (policy statement No.3) as you correctly pointed out " valves qualified for closure (emphasis added) at design pressure or peak LOCA pressure." Therefore, this may be a significant backfit if the current valve arrangement requires major changes, which I am not able to address at this time.
- With regard to the capability of vent pipes to channel core debris to the suppression pool, additional questions should be asked whether and how long
- can the downcomer pipes in Mark II's be counted on to maintain their structural capability. The wall thickness is not very large. Therefore, some point in time the hot core debris may burn thru the downcomer pipes and create a bypass, resulting in overpressurization of the wetwell, which is designed for 45 psig.
4 Unle'ss, we miss the exact overpressure / temperature sequence, I believe it is not as simple as we think to modify the current system for coping with the i'
Contact:
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severe accidents of a size like Chernobyl. Therefore, I agree with your opening statement that the containment performance may lend itself to generic treatment, possibly by a rulemaking because it is not clear in my mind which one is more vulnerable.
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Engineering Branch Division of BWR Licensing cc w/ enclosure:
G D.Lainasiisin M. W. Hcdges C. P. Tan J. Lombarac
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t PERFORMANCE OF BWR CONTAINMENTS IN SEVERE ACCIDENTS
1.0 INTRODUCTION
i
! On the basis of geometrical configurations, containments for BWR plants can be classified into Mark I, Mark II and Mark III, that may be constructed of
- steel with concrete shielding or of concrete (reinforced or prestressed) with a steel liner. Therefore, generally, there are six types of BWR containments j summarized as follows
Confiouration Materials of contruction No. Design Pressure Psig Mark I steel 23 62 concrete 2 Mark II steel 1 45 concrete 8 j Mark II; steel 4 15 j concrete 4 The ability of each of these containments to cope with severe accidents i depends to some extent on the type of containments to which it is categorized.
The geometrical configurations of the Mark I, II, and III containments are
- snown in Figure 1. The scenario of severe accident generally involves high i
pressure and high temperature accompanied with or without an explosion. Mark
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I and Mark II containments are inerted with nitrogen and Mark III containments are installed with igniters. Therefore, we believe hydrogen detonation is j highly unlikely. Even though Mark I and Mark II containments have not been evaluated for their ultimate capacity to resist pressure loading, it can be concluded from the results of ultimate capacity evaluations for Mark III and PWR containments that all the containments, either steel or concrete, have the capability to resist pressures two to three times the design pressure.
3 One commonality of all types of Mark containments is that they are enclosed in a reactor building which can attenuate the release of radioactive fission j products.
i 2.0 EVALUATION OF CONTAINMENT STRUCTURES i
i In the following each type of containment is examined in some more detail for its capability to cope with all aspects of a severe accident:
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2.1 Mark I Steel Containment The Mark I steel containment consists of a bulb shaped drywell and a torus suppression pool of circular cross-section with vent pipes connecting the former to the later. Both the drywell and the torus suppression pool are enclosed in a reinforced concrete structure. In case of a severe accident involving core melt-down, the structural integrity of the steel torus will be challenged first by high pressure which may reach to a level to rupture the torus and finally the reinforced concrete shielding structure.
This will result in an uncontrolled release of radioactive materials.
However, as indicated in section 3.0 valves are installed on the drywell and the torus and the exhaust gas is filtered and vented through plant stack into the atmosphere. Therefore, the release would be controlled, if the logics of the valve operation can be changes to ensure operability to open at design or higher pressure.
If the reactor melt-down proceeds to the extent that core debris falls down on the drywell floor, the molten material will attack first the concrete fill above the bottom of the drywell bulb and will penetrate the drywell steel shell and the foundation concrete. Some of the molten material will flow through the vent pipe into the suppression pool resulting in the rupture of the torus shell and its concrete foundation. The postulation of such a scenario means the reach of eventuality of uncontrolled release of radioactive materials. If the reactor accident reaches to such a stage, there is not much .
difference in the consequence of radioactive material release regardless which containment system is used.
2.2 Mark I Concrete Containment Mark I concrete containment has basically the same geometrical configuration as Mark I steel containment. This type of containment is attained by covering the inside surface of the concrete structure with a steel liner so as to make it leaktight and by providing enough reinforcing steel 50 as to enable it to resist the design loads with sufficient margin of safety. It acts as a pressure retaining structure as well as a shield. Basically there is not much difference between Mark I concrete containment and Mark I steel contain-ment venting of the drywell and the torus is the same as that described for the steel containment.
2.3 Mark II Steel Containment The geometrical configuration of Mark II containment is shown in Fig. 1.
Basically it consists of a drywell on top of a suppression pool with a diaphragm separating the two. The primary containment consists of a steel pres'sure-retaining vessel enclosed by a concrete shield wall both supported by a concrete base mat. There is a small gap (1" to 2") between the steel vessel and the shield wall. Because of the proximity of the steel vessel to the shielding structure. The shielding structure will be subjected to the
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c pressure loading transmitted through the steel vessel as the internal pressure increases and the deflection of the steel vessel is such that the gap between j the two structures is closed. As a result, the capability of the steel vessel to resist internal pressure is enhanced. In order to control the release of radioactive materials, pressure relieve valves can be installed on the steel vessel by passing through the shield building and the filters. The contents from the containment after filtered are released into the reactor building and i can be filtered (or plate-out) again and released into the atmosphere.
2.4 Mark II Concrete Containment
, Basically there is not much difference between the steel containment and the concrete containment. The latter employs a steel liner to maintain its -
- leaktightness. The pressure resistance capacity depends on the reinforcing
- or prestressing steel. The use of pressure relieve valve and filter is nearly the same as for steel containment.
2.5 Mark III Steel Containment This containment has the same configuration as a PWR dry containment. The steel containment is enclosed in a shield building. An annulus with a width of 3 to 5 feet exists between the two structures. In view of this feature, the containment contents with radioactive material can be released into this annulus where after being filtered can be released into the reactor building, which in turn, will release into the atmosphere. To attain such a process of releasing pressure, relieve valves can be installed on the steel containment and the shield building.
2.6 Mark III Concrete Containment
, For the concrete containment there is no longer the existence of the annulus
' as in the case of steel containment. The features of relieving the radioactive materials and filtering are basically the same as for Mark I, II i
concrete containments.
- 3. 0 BWR PURGE AND VENT CCNFIGURATION The containment inerting system maintains an atmosphere of nitrogen gas within I
the drywell and torus air space during power operation, and performs two functions, purging and makeup. In each purge supply line the two valves in series are normally closed during reactor operation to isolate the containment vessel from the purge line.
Air for purging the drywell is supplied from the reactor building ventilation supply system through 18-inch air operated butterfly valves, remotely, controlled and interlocked to close automatically on containment isolation i signal. These valves are normally closed during reactor operation.
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Air for purging the torus is supplied directly from the reactor building atmosphere when a slight vacuum is established by exhausting the torus to the stack, and is controlled by the torus vacuum breaker valves. These valves are automatically opened by differential pressure switches to prevent excessive vacuum in the torus and will also open on loss of air pressure. The valves are normally closed with positive or atmosphere pressure in the torus.
Purge exhaust from the containment is drawn to the stack through either the reactor building vent exhaust fan or the SGTS fan, depending on the radio-active levels of the exhaust gas. When the exhaust is vented directly to the stack without gas treatment, torus flow is through the 12-inch valves, and from the drywell to the stack. Without gas treatment, the flow is through the 18-inch valves. These valves are air operated butterfly valves, remotely controlled and normally closed. Air accumulators are provided as a backup air source to ensure closure on a loss of control air signal, and the valves are interlocked to close automatically on high drywell pressure or low-low reactor water level.
When the exhaust is routed through the SGTS, the 18-inch valves are left closed and the 2-inch bypass valves are open, limiting the flow to less than 2600 CFM and reducing the drywell pressure to prevent damage to the SGTS filters.
When the torus pressure exceeds drywell pressure by 0.5 psi the drywell vacuum
- breaker check valves automatically open to equalize pressure.
When the valves are used for venting, they are manually opened when the drywell pressure is greater than 1.5 psig, but less than 5 psig. They vent either to the normal building exhaust fan or to the SGT5. The vent connection accomodates pressure increases during normal drywell heatup at station startup, and are interlocked with mode switch position and containment isolation requirements.
4.0 CONCLUSION
From the above it can be stated that the capability of BWR containments in case of severe accidents can be greatly enharced by venting and filtering.
The purpose of such a process is to avoid burst-mode failure of the containment and as a result the release of radioactive material will be orderly. Under accident conditions the preferred venting is through the torus (per Emergency Operating Procedures, E0P's). Under severe accident conditions the drywell vent valves can also be used to protect the primary containment from over-pressure.
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