ML19291C038
| ML19291C038 | |
| Person / Time | |
|---|---|
| Issue date: | 02/14/1978 |
| From: | Siegel B Office of Nuclear Reactor Regulation |
| To: | Fitzpatrick R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19291C039 | List: |
| References | |
| REF-GTECI-A-30, REF-GTECI-A-44, REF-GTECI-EL, TASK-A-44, TASK-OR NUDOCS 8001110122 | |
| Download: ML19291C038 (4) | |
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FEB 1 4 '973 MEMORANDUM FOR:
Robert G. Fitzpatrick, Power Systems Branch, DSS V
THRU:
Thomas M. Novak, Chief, Reactor Systems Br., DSS 7' Gerald Mazetis, Section Leader, Reactor Sys. Br., DSS FROM:
Byron Siegel, Reactor Systems Branch, DSS
SUBJECT:
CATEGORY "A" TAR-A-30, LOSS OF AC/DC POWER As you know, PSB has requested RSB assistance for two tasks identified in TAP A-30.
We had scheduled "around" this Category A activity as requested in the Task Action Plan. This scheduled period is now complete and I am continuing my other scheduled activities. As was discussed with you previously on several occasions, we should recognize that due to the nature of the RSB assigned task in A-30, the need exists for task group meetings to understand the ground rules, such as the scope, depth, and interfaces associated with this study area.
I again suggest such a meeting prior to further work on this activity.
Some RSB work was performed on Task 1 (see the enclosed summary).
The remaining work will await your direction; however, it must be realized that decisions on RSB involvement will be made on the basis of an evaluation of my other work activities at the time.
imt B9ronSieel Reactor Systems Branch Division of Systems Safety
Enclosure:
As Stated cc:
D. Ross
- 8. Siegel R. Tedesco G. Mazetis T. Novak RSB Members F. Rosa 1734 249 B. Sheron
Contact:
B. Siegel, NRR x27341 8 0 0111012 E
ENCLOSURE Task No.1 - Provide assistance to Power Systems Branch for definition and scoping of the concern.
This will include definition of the safe plant condition which must be achieved to terminate the postulated sequence of events including identification of those systems which must be opera tional.
A " safe shutdown condition" has been interpreted in several ways in the past, depending on its application.
However, the definition judged to be applicable to this situation is given in 10 CFR 50.46 which states that "the calculated reactor core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core." Maintaining the core at an acceptably low temperature is interpreted as that below which significant fuel rod failures or degrada-tion cannot occur and a coolable geometry can be maintained. This definition does not require distinguishing between a hot and cold shutdown condition in the short term following a loss of all AC/DC power.
As stated in the Brian Sheron memorandum dated January 25, 1978, the time available before the core uncovers below the midplane, assuming no ex-tornal cooling source, varied from approximately one hour to upwards of four hours depending on the type of plant (BWR or PWR) and the plant design (B&W or Westinghouse). To extend the time period before restoration of power, water from an external source would have to be provided to maintain a coolable geometry. This water would have to be pumped into the reactor vessel for a BWR or into the steam generator of a PWR, Doth of which would be at or near thei'r pressure relief setpoints.
There are two high pressure systems that can be potentially operational, the RCIC system for BWRs and the auxiliary feedwater system for PWRs.
Both of these systems utilize steam turbines to drive high pressure water pumps.
In the RCIC system, steam is tapped off the main steam lines between the reactor vessel and the main steam line isolation valves.
The water supply to the pump would be from the condensate storage tank with the suppression pool providing a backup water source.
The water is pumped into the reactor vessel.
For the PWR auxiliary feedwater system, the steam is tapped off the main steam line upstream of the main steam line isolation valves. The water is supplied to the pump from the condensate storage tank and pumped into the steam generators.
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_2-In general, the turbines and valves for both these systems are remotely controlled during normal operation and require DC power.
For BWRs the turbines, which have DC speed controllers, can be manually operated and controlled. This also is true for most PWRs; however, a few may have mechanical hydraulic turbine speed controllers, which do not require DC power. Some of the earlier PWRs may also have pump and
.y turbine bearings that are cooled from an externally powered AC source, in which case they would become inoperable. Operation of either of 4
these systems assumes that the valves which have to be open are e
manually operable. The cpity and capability of these systems to p
provide sufficient makeup water to maintain a coolable geometry will }p c5 g'
have to be determined on a plant-specific basis (Task 4).
Following reactor trip, the decay heat and amount of steam blowdown may be I
T greater than the makeup capacity of these systems.
However, with the decrease in decay heat, these systems will provide more makeup capacity than the quantity of steam released through the relief valves (primary side on BWRs and secondary side on PWRs). Although utilization of thee systems prolongs the time available to restore power and reduce p
system pressure, cold shutdown Cannote achieved in this mode of operation. AC and DC power wi n eventually be required to permit operation of the ECCS and/or RHR systems to completely depressurize the reactor and achieve a cold shutdown condition.
A complete evaluation of this event would require a determination of the time it would take for the RCIC and auxiliary feedwater systems to become operational under the adverse plant conditions expected without AC or DC power. This would depend upon the availability of plant telecommunications systems, locations of emergency DC lighting, avail-ability of experienced personnel to manually operate these systems, equipment Ovailability (i.e., portable tachometer to determine turbine rpm),
and the capability of the operator to assess plant conditions to initiate the proper action. Since most,if not all, instrumentation is not available to him, this could pose a serious probler.
Also, it is important to emphasize the criterion to be used in defining the coolant level in the core at whi h significant fuel damage and loss of a coolable geometry might occur (ll. The Brian Sheron memo implies that perhaps this criterion could be at the point where the vessel water level reached the core midplane. However, additional justification to support this criterion would need to ba documented.
Clearly, a more conservative and perhaps desirable position in the interim would be to prevent core uncovery.
The disadvantage to this approach is that the time required to achieve operability of the RCIC for a BWR (or the auxiliary feedwater system for a PWR) would be more limiting.
U)It should be clear that the amount of fuel damage acceptable will be dependent on the probability of occurrance of this event, which is to be evaluated under another part of this task action plan.
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I A final consideration would be the required duration of operability of the RCIC or auxiliary feedwater which is dependent upon the assessment of the time required to restore power.
If the time to restore power is too long, some plants may require modifications or alternate means of providing power within an acceptable time limit.
In summary, it appears that for most plants a system is available to provide a standby source of water which will prevent or prolong core uncovery and maintain the core in a safe shutdown condition.
These systems require no AC or DC power but would have to be manually operated under adverse conditions.
Restoration of power is still required to bring the plant to a cold shutdown condition, howevcr, these systems will prolong the time available.
There are many interrelated considerations that will have to be resolved to determine the availability and the effectiveness of these systems in bringing the reactor to a safe shutdown condition (Task 4).
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