ML20206L931
| ML20206L931 | |
| Person / Time | |
|---|---|
| Issue date: | 08/09/1986 |
| From: | Thadani M Office of Nuclear Reactor Regulation |
| To: | Bernero R, Houston R, Lainas G Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8608200411 | |
| Download: ML20206L931 (10) | |
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AUG 0Y i*
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MEMORANDUM FOR: Robert M. Bernero, Director Division of BWR Licensing i
R. Wayne Houston, Deputy Director l
Division of BWR Licensing i
Gus C. Lainas, Assistant Director i
i Division of BWR Licensing i
FROM:
Mohan C. Thadani, Project Manager i
BWR Project Directorate #2 Division of BWR Licensing i
SUBJECT:
GENERIC REQUIREMENTS FOR BWR CONTAINHENT RESPONSE TO SEVERE ACCIDENTS Attached for your' review is a preliminary draft of our evaluation supporting
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the proposed advance notice of Generic Letter for BWR Containment requirements.
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The evaluation was prepared by Wayne Hodges and Jerry Hulman.
Please send your comments to me by Thursday, August 14, 1986.,
Driginal signed by j!
Mohan C. Thadani, Project Manager
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BWR Project Directorate #2 Division of BWR Licensing l
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BWR MARK I & II CONTAINMENT PERFORMANCE i
DURING SEVERE ACCIDENTS i
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1.0 INTRODUCTION
I In its Severe Accident Policy Statement of August 8,1985, the Comission indicated its objective is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an j
accident should one occur. Further, tht: Connission stated that the examination of individual reactor risks will " include specific attention to containment j
performance". The examination of severe accident risks at BWR plants with Mark l
I & II containments has lead the staff to conclude that a number of significant j
improvementsincontainmentperformance(inbothdesignandoperation)canbe made at moderate to low cost.
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2.0 CONTAINMENT PERFORMANCE ASSESSMENT l
Containments are required to protect the public from the consequences of
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accidents. The design and sizing of containment systems are based on the j
pressure and temperature conditions which result from release of the reactor coolant in the event of a design basis loss-of-coolant accident (LOCA). The containment design basis includes the effects of stored ene'rgy in the reactor coolant system decay energy, and energy from metal-water reactions including i
the recombination of hydrogen and oxygen. The containment system is not j
required to be a complete and independent safeguard against a design basis LOCA by itself, but functions to contain any fission products released while the emergency core cooling system cools the reactor core.
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Notwithstanding that a postulated design basis LOCA is not expected to produce l
more than a few percent fuel failures, an accident radiological " source term"
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is used in calculating offsite dose consequences representative of a substantial core melt accident (10 CFR 100). Even for this source tenn,
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containments are designed such that calculated offsite doses are unlikely to result in a significant early or latent health hazard if the containment were i
to maintain its low leakage capability. What is at issue is the capability of j
containments to perform a mitigating safety function as long as practicable l
during very low probability beyond design severe accidents when the i
consequences and risks due to containment failure may be very significant.
This significance is amplified for two BWR containment types (MK I & II) where, notwithstanding the positive pressure suppression feature, they are relatively small, and their likelihood of failure in a severe accident is judged to be i
j higher than the larger MK III pressure suppression or " dry-type" containments, i
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Overall plant core melt probabilities for MK Is and IIs have been estimated at
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values of from one in a thousand per reactor year to less than six in a million for eight BWR designs. Contemporary analyses break down such probabilities into classes and subclasses of accidents.
IDCOR has proposed five c' lasses of i
events for BWR core melt accidents, depending on the initiating event and j
containment response as follows:
loss of core cooling with containment at low pressure and failure after core melt; loss of core cooling with containment failure before core melt; t
l loss of core cooling with containment failure soon after core melt due to high containment pressure at the time of core melt;-
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loss of core cooling with containment failure before core melt due to l
failure to depressurize; and j
containment bypass.
The staff concludes that BWR core melt accidents can be grouped into just three categories as follows:
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transients within which ATWS is most important; losses of heat removal capability in eitaer the vessel or containment within which station blackout sequences dominate; and bypass sequences.
The sum of the core melt probabilities for all classes and subclasses of accidents is considered to be the overall core melt probability. Our review of these probability estimates to date generally indicates they are low, but given a core melt, the estimates of likelihood of Mark I and II containment failure have been high relative to other containment types *.
In all of these past evaluations little or no credit has been given to the ability of operators to prevent or mitigate such accidents.
Furthermore, little or no credit has been given to features which can be used with relatively modest effort to prevent or mitigate accidents. For most accidents considered, the core is postulated to melt; interact with steam, water, and the structural features in the vessel and coolant system; melt through the vessel; and attack the concrete and structural features of the lower containment. Depending on the sequence of events, the
- The Reactor Safety Study (WASH-1400, NUREG-75/14) indicate's a conditional containment failure probability for the BWR Mark I containment reference plant (peach Bottom) of about 90% (Table 5-3, page 81)).
That is, given a core melt in a BWR with a Mark I containment (Peach Bottom there is a 90% chance of containment failure.
In the November 1984 IDCOR Technical Sumary Report, Nuclear Power Plant Response to Severe Accidents, the estimate for Peach Botton was about 20% (Table 10-1, page 10-6). For a BWR Mark II containment (Limerick), BNL estimated almost a 100% likelihood of containment failure given a core melt (BNL 33835, April 1904).
In only the IDCOR evaluation was containment venting considered.
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4 containment is postulated to fail either before or after vessel melt through.
For the remainder of the accidents postulated, the containment would be bypassed, allowing radioactivity a direct path to portions of a plant not capable of containing the releases, but with some capability to attenuate radioactivity.
Mark I and II containments respond to heatup of the fuel in the vessel directly or indirectly. The direct transfer of energy is through pipe breaks, or through blowdown into the suppression pool.
Indirectly, radiant heat is transferred through the vessel and piping. The blowdown or depressurization, if it works, and the use of the relatively large quantity of suppression pool 2
water as a heat sink and fission product scrubbing device, will work in combination with the structural capability of the containment (including penetrations) to mitigate the high temperatures, pressures and radioactivity released in a core melt. Core melt scenarios have been identified which can l
produce conditions in the containment that can lead to failure. However, there is strong evidence that containments are capable of withstanding substantially higher pressures; evidence that can be capitalized on to provide additional protection to the public at modest to low cost. The longer a containment can j
be expected to hold, the greater the likelihood that failure can be avoided, but if failure were to occur, greater reduction in radioactivity released can be achieved. Actions that can be taken to prevent a catastrophic failure of containment before the fission product attenuation advan'tages may be fulfilled include such items as operator actions to vent the wetwell space above the suppression pool, providing reliable spray capability, and quenching core debris. Spraying the core may also prevent a core melt from occurring.
In a core melt accident with temperatures in excess of 5000 degrees F, fission I
products are released from the fuel in three general groups. The noble gases and the more, volatile species of fission products are released from the fuel i
relatively early in a core melt upon the occurrence of fuel melting in the vessel. Later, the less volatile species are released as the fuel melts down l
into the vessel and combines with the in-vessel structural materials.
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Finally, after melting through the vessel, the refractory fission products (suchasplutoniuminrelativelysmallquantities)arereleasedthrough l
interactions of corium with concrete on the floor of the containment.
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l The amount of radioactivity that could be released to the environment in core melt or degraded core accidents has been the subject of considerable debate for I
a number of years. Present estimates (NUREG-0956) for MK I and II BWRs indicate that substantial quantities of important fission products can be j
released in a core melt accident, but can be reduced by a number of actions to enhance containment performance. Within the core of a BWR MK I or II at full j
power are over five billion curies of radioactivity. Severe accident releases J
to the environment for a MK I or II have been estimated to exceed 40% of such relatively important fission products as iodine and cesium (releases of over l
300 million curies of iodine and over seven million curies of cesium for a j
3458 MWt reactor). Backfits specifically designed to improve BWR MK 1 and II I
containment performance have been evaluated. Those backfits that have been justified are discussed in Section 3, and are summarized in Section 4.
3.0 MITIGATION FEATURES CONSIDERED i
The vulnerability of MK I and MK II containments to failures due to severe l
accidents and the source term attenuation capability'of the containments can be j
greatly enhanced by a few cost effective modifications to the containments and support systems. Although some of the proposed changes could provide 1
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additional core cooling capability and thus further reduce the low probability r
of core melt, the emphasis here is on mitigation of a severe accident should l
one occur. The goal is to obtain substantial assurance that, given a core j
melt, the likelihood that the containment will be breached is substantially j
reduced, and the release of fission products to the environment will be minimized. This goal can be achieved with the simple modifications discussed below:
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Assure wetwell venting. Although containment venting has been recognized i
by the BWR owners in the Emergency Procedure Guidelines as an effective i
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6 pressure control measure, venting is not practical today at many plants because of limitations on ductwork, isolation interlo b and valve accessibility. These limitations are readily correctable.
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Make containment sprays independent of normal power and water sources.
This can" easily be obtained at most BWR plants by the addition of interties between the RHR system and the nornal fire protection systems.
These fire pumps are typically powered by dedicated diesels and have a variety of possible water s'ources. Connections for fire hoses on the exterior of containments and use of nearby pumper treiks would also be useful.
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Minimize deinerted time. Typically, plants are allowed to be deinerted 24-hours after startup and 24-hours prior to shutdown. Minimization of this deinerted time would reduce the vulnerability to hydrogen combustion.
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For Mark I containments add masonry structure beneath torus. Such structures would retain the pool water and core debris should the debris cause failure of the torus.
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Improved praeuures.
Implementation of the most recent symptomatic emergency procedure guidelines would remove uncertainty as to how to cort with a severe accident and, greatly-improve the probability of successful mitigation. The implementation should also substantially improve operator response to accidents.
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Assured automatic depressurization capability.
Installation of a few dedicated batteries would assure that, if core melt should occur, it could occur at low pressure. Analyses have shown that a core melt at low pressure provides a less serious threat to the containment than does a high pressure core melt.
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7 4.0 DISCUSSION OF BENEFITS Most Mk I and Mk II BWRs, as currently configured, exhibit a relatively low probability of core melt. However, given a core melt, the fragility of the containment is such that a significant release of fission products to the environment is likely. The challenge to containment integrity comes not from design basis accidents, but from accidents involving multiple failures such as station blackout for an extended period, or from anticipated transients without scram (ATWS).
Tests and analyses have shown that the containments have considerable conservatism for coping with design basis accidents. This extra design basis margin can be used to extend the capability of MK I and Mk II containments to cope with core melt accidents at moderate cost. Some of the modifications proposed would have the added advantage of further reducing core melt likelihood.
For a core melt accident which may lead to containment. failure by overpressure, such as for a severe ATWS, containment pressure control is very important. The three primary means of supplementing the suppression pool for pressure control are wetwell spray, drywell spray and containment venting. These pressure control methods are all recognized by the BWR Owners as potentially effective and are explicitly called out in the emergency procedure gu'idelines developed by the BWR Owners' Group.
Fortunately, although these actions cannot always be accomplished in existing plants because of design inadequacies, the deficiencies are easily corrected.
Containment venting is impractical at many plants only because the ductwork in the reactor building was not designed for venting loads and the containment vents are not remotely operable. Under design basis conditions, there is no need to vent'and the vent valves are designed to close and stay closed on loss of motive power (air or electric). Containment isolation interlocks must be
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defeated for the vents to open under accident conditions. At modest cost, the vent valves could be made remotely operable by installing backup nitrogen bottles for opening the valves and by installing a bypass, controllable from the control room, for the isolation interlock. The ductwork could be modified by installing a tee and valves to vent through piping directly through the roof of the reactor building.
Indeed, such vent pipes already exist at some BWRs.
There would be no need to replace the bulk of the duct work. By using vents which are connected to the wetwell, fission products would be scrubbed in the l
suppression pool, thereby minimizing the release of fission products. Venting can arrest or substantially delay containment failure. Studies at BNL have shown that venting can reduce the fission product release by more than an order of magnitude for slowly developing events such as station blackout or ATWS.
The additional time to containment failure (without operator action) greatly enhances the ability of the operator to prevent containment failure.
Although venting would be very effective in extending containment integrity, less drastic measures are preferred, if possible.
Except for the ATWS case, containment heatup and pressurization can be controlled with relatively modest flow rates for wetwell and drywell sprays. These sprays have the added advantage of washing surfaces and scrubbing fission products in the containment atmosphere; thus, fission product release would be further reduced.
In addition, drywell spray would be useful in cooling the molten core debris if it were spreading across the containment floor. Flow rates on the order of 200-400 GPM would be adequate for decay heat removal provided a cool source of water is available. These rates could be obtained with fire systems or a fire truck and an external hose connection on the containment. Accessibility of 1
valves and connection locations during an accident would have to be assured.
An appropriate source of water would be the condensate storage tank, a lake, pond or nearby stream. Larger flow rates may be needed to obtain spray flow rather than dribble flow from the spray nozzles.
Because the wetwell and drywell sprays are parts of the RHR system, the added systems would further reduce the probability of core melt.
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9 Because of the small Mk I and Mk II containment volumes, the combustion of hydrogen and oxygen poses a threat to containment integrity. These containments are nomally nitrogen inerted during operation to preclude combustion. However, the containments are permitted to be de-inerted for 24-hour periods after startup and prior to shutdown. Although this minimizes outage time, it increases containment vulnerability during transient periods.
Minimization of these allowed de-inerted periods and assuring the availability of hydrogen recombiners would improve combustible gas control.
Should the molten core debris reach the torus and melt through, then a masonry structure beneath the torus would retain sufficient water from the pool and melt material, thus preventing release outside the reactor building. This structure, in some plants, would be r.o more than a dam at the doorway to each corner room. Such a structure could be and easily built at very modest cost.
Not all utilities have fully implemented the latest emergency procedure guidelines (EPGs). The BWR Owners' Group has recently completed Revision 4 to the EPGs and the staff has started its review. We believe these EPGs to be very effective in responding to severe accidents. We, therefore, rccc=cnd that the implementation of Rev. 4 to the EPGs be expedited.
If a core melt should occur, the impact on the containment is less severe if the reactor vessel is depressurized at the time that the core debris melts through the vessel. Therefore, the installation of a dedicated DC power supply to assure operability of the automatic depressurization system would assure a low pressure melt-through.
It would also be useful in preventing core melt in some cases. Because DC power would only be needed for a few solenoid valves, the cost should be minimal and would likely involve only a few batteries.
Two rules influence severe accident risks at BWR MK I & IIs; ATWS (50.62) and the proposed" Station Blackout Rule. Accelerated implementation of both would reduce the risk exposure at BWR MK I & IIs.