ML20002C552
| ML20002C552 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/07/1978 |
| From: | Skibitsky W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8101100465 | |
| Download: ML20002C552 (3) | |
Text
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[][ T Director, Nuclear Reactor Regulation Att: Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission fy.7
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Q'.A il&%.y DOCKET 50-155 - LICEMSE DPR BIG ROCK POINT g.y Ni e.d?
PLA:~T - RESPONSE TO LETTER DATED FEBRUARY 1, W7
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1978: HIGH LEVEL FEED PUMP TRIP By letter dated February 1, 1978, Consumers Pcver Company was requested to provide the staff with a schedule for installation of an automatic reactor feed-water pu=p trip based upon high reactor vessel water level for the Big Rock Point Plant. Upon evaluation of the necessity and desirability of such a modification, Consumers Power Company has concluded that the installation of
- his trip is inappropriate and unnecessary for Big Rock Point. This evaluation is based upon the fellowing censiderations:
(1) The high reliability of the feed-water control system, (2) the existence of a steam drum at Big Rock Point and the large feed volume it affords, and (3) the' unusually hish availability required of the reactor feed-water system under specific LOCA conditions.
The Big Rock Point feed-water control system has operated reliably for 15 years with no known problems relating to inadvertent flooding of the primary steam drum.
In general, the feed-water control system is a three-element contrclier utilizing steam flow, feed flow, and steam drus water level signals.
The system is designed to maintain drum level within i 1 inch of progranted level during steady state operation, and to handle all no mal plant load svings without re-sulting in reactor trip on 1cv drum level (8.5 inches below normal level). Stec=
flow is the primary element in the controller. A mismatch between steam flow and feed flov is anticipatory of an i=pending drum level deviation and vill result in apprcpriate controller action. For example, a step increase in steam flow, and the resulting reduction in drum pressure, causes an immediate swelling of the drum level due to flashing.
The centroller, however, vill cause an increase in feed-water flow in anticipation of the eventual fall in drum level ss the primary systen fluid inventory is depleted based upon the steam flow / feed flev misratch.
In the unlikely event of a large reduction in steam flow (ie, caused by a turbine trip without bypass, for example) the drum level vould rapidly fall due to the collapse of voids in the primary system. The operation of the centroller vould be to initially reduce feed-water flow in response to the high steam flev/ feed flow mismatch and thus avoid overfilling of the drum. The controller would then
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continue to supply some water to the drum until normal level was reached.
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'Due to.the high free volume of the primary steam drwn,' the potential for -
. completely filling the drum and overpressurizing the primary system is remote.
The steam drum'which contains the ' steam separators and dryers, as well as. the :
feed-water spargers,' has ' a free volume of about '1,100 cf.
During normal oper-atien the drum is half full. Failure of the control system could result in-filling of the drum beyond the normal water level. Assuming such a failure,
'high drum level alarms vould be initiated.at h" and 13" above dru= sidplane.
The alarm at -13" is part of the reactor. depressurization system, 'is four-channel
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redundant, and is environmentally. qualified. Under the worst conditions, with the reactor tripped and assuming a very high ' feed-water. flow rate of 2,200 gpm,-
a the operation would have at least 2.4 minutes after the first alarm to terminate' the transient before the drum would fill. However, if the amount of available condensate is considered, the feed pumps can be shown to trip on lov suction.
I pressure' before the drum fills. For other cases, the drum would fill more slowly, thus allowing adequate time for operator action to terminate. the transient.
It should be noted that aside from the primary safety valves, no safety-related equipment or equipment required for. the ' orderly shutdown of the reactor would be affected by the filling of the steam drum and that the filling of the drum cannot
- inhibit initiation of any required safety features. A solid drum condition vould inhibit the actuation of the reactor depressurization system (required for small'
- break LOCAs), however, the availability of feedvater can only improve the conse-quences of the LOCA (ie, core uncovery is not possible if water remains in the steam drum). Filling of the drum may result in damage to the primary safety
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relief valves..However, as noted above', there vill exist adequate time after the j; ~
high drum' level alarm is actuated before operator action is required to terminate the level rise. Thus, the possibilities of this even occurring is considered remote.
One other aspect of reactor feed-water system performance must be addressed. For operating Cycle 15, Big Rock Point is exempt-from the equirements of 10 CFR-50.h6 and Appendix K, when applied to a LOCA caused Fy a break in the redundant corefspray line. This exemption was granted, in part, based on the reliability displayed by the reactor feed-water system both in operation and capacity.
Clearly, any modification to install additional trips to the reactor feed-water pumps would only serve to lessen the overall reliability of the system to perform under LOCA conditions and, therefore, would be an unnecessary risk.
In su= mary, based upon the proven reliability of the feed-water control system, the excess capacity of the steam drum when compared to the normal feed-water flev i
rate, and the required availability of the reactor feed-water system for the un-likely event of a LOCA caused by a break in the redundant core spray line, Consu=ers Power Company concludes that the installation of an automatic high
. reactor water level trip for the reactor feed-water pu=ps is an unnecessary and undesirable modification and, therefore, vill not be made.
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William S Skibitsky
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Senior Licensing Engineer CC: 'JGKeppler, USNRC
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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
DISTRIBUTION FOR INCOMING MATERIAL 50-155
' REC: ZIEMANN D L ORG: SKIBIT9KY W S DOCDATE: 03/07/78 NRC CONSUMERS PWR DATE RCVD: 03/10/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 0 RESPONSE TO LTR DTD FEB.
1, 1978 RE: HIGH LEVEL FEED PUMP TRIP.
' PLANT NAME: BIG ROCK PT REVIEWER INITI AL:
XRS DISTRIBUTOR INITIAL:p l 0000000**********
DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS ******************
GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE AOO1)
FOR ACTION:
BR CHIEF J _
T ONLY(7) n b EG FILE **LTR ONLY @
E NRC PDR**LTR ONLY(1)
INTERNAL:
I & E** N 4LY(27 OELD**LTR ONLY(1)
HANAUER**LTR ONLY(1)
CHECK **LTR ONLY(1)
EISENHUT**LTR ONLY(1)
SHAO**LTR ONLY(1)
BAER**LTR ONLY(1)
BUTLER **LTR ONLY(1)
GRIMES **LTR ONLY(1)
J COLLINS **LTR ONLY(1)
J.
MCGOUGH+*LTR ONLY(1)
EXTERNAL:
LPDR'S CHARLEVOIX, MI**LTR ONLY(1)
TIC **LTR ONLY(1)
NSIC**LTR ONLY(1)
ACRS CAT B**LTR ONLY(16) l DISTRIBUTION:
LTR 40 ENCL 0 CONTROL NBR:
7606~.~ 0125 SIZE: 2P 00000n*****************************
THE END
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P00R ORIGINAL
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