ML20246B681
| ML20246B681 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/30/1989 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-88-17, NUDOCS 8907100056 | |
| Download: ML20246B681 (5) | |
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E BALTIMORE GAS AND ELECTRIC CHARLES CENTER.P.O. BOX 1475 BALTIMORE, MARYLAND 21203 GEORGE C. CREEL vier PRESIDENT Nucttaa EntRoy -
(300 aso-44 ss '
June 30,1989 9
'U. S. Nuclear Regulatory Commission.
'. Washington, DC 20555 ATTENTION:
Document Control Desk -
SUBJECT:
. Calvert Cliffs Nuclear Power Plant Unit Nos.1 &.2; Docket Nos. 50-317 &.50-318 Response to Generic Letter 88-17. Loss of Decay Heat Removal"
REFERENCES:
(a) Letter from. Mr. J. A. Tiernan (BG&E) to NRC Document - Control Desk,' dated January 3,1989, same subject (b)' Letter from Mr. J. A. Tiernan (BG&E) to NRC Document Control Desk, dated January 27, 1989, same subject
. Gentlemen:
In response to Generic Letter 88-17, Baltimore Gas and Electric Company implemented the expeditious actions described in. Reference (a). Our long-term program enhancements, which do not include. hardware modifications, are detailed. in Reference (b). The purpose of this letter is to provide the planned modifications' to our Reactor Coolant System level indication, as well as our actions taken to address Shutdown Cooling
. Instrumentation concerns. These are both described. in Enclosure (1).
' Should you have any questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours, l$
00 GCC/LSL/ dim g[f Enclosure l
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i 8907100056 890630 PDR ADOCK 05000317 P
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Document Control Desk June 30,1989 Page 2 cc:
D. A. Brune, Esquire
,J.
E.
Silberg, Esquire R. A.Capra, NRC S. A.McNeil,NRC '
W. T. Russell, NRC II. Eichenholz/V. L. Pritchett, NRC T. Magette, DNR
ENCLQSURE I CALVERT CLIFFS NUCLEAR POWER PLANT FOLLOW-UP RESPONSE TO GENERIC LETTER 88-17 Our letter of January 27, 1989 responded to Generic Letter 88-17 and outlined our program enhanecments for operation during reduced inventory conditions. In that letter, we advised that we were reviewing our present Shutdown Cooling (SDC) System and tne Reactor Coolant System (RCS) arrangements to determine the best parameters to monitor system performance. We also stated that we would identify the instrumentation which we will add or modify to monitor RCS level and SDC system performance.
The purpose of this subsequent response is to describe:
o Actions that are planned and will be executed at Calvert Cliffs to address the Generic Letter's concern of whether there is adequate instrumentation, with a sound technical basis, to monitor the performance of the SDC system, and o
The planned modifications to improve our indication of RCS level at reduced inventory conditions.
Briefly, our project team performed a thorough review of operating experience at Calvert Cliffs and other plants, reviewed installed instrumentation, conducted a
literature search of technical topics associated with vortexing, air binding and pump cavitation, and met with the SDC pump manufacturer to review the pump design. While these inputs were very useful in validating the key parameters to be considered, it is desirable to obtain more plant-specific data to establish the most useful parameters to monitor SDC system performance.
The parameters considered as potentially good indicators of SDC system performance include pump acoustics and vibration, pump suction pressure, pump motor current, and pump discharge pressure, in addition to RCS level, water temperature, and pump flow.
In order to perform a thorough evaluation of these parameters as indicators of SDC system performance, it was concluded that we should conduct SDC system performance tests. These tests are presently planned for Unit 2 during the current refueling outage.
The above parameters will be monitored using installed instrumentation, as well as temporarily installed instruments that permit automatic data recording. Data will be taken at several water levels from the top of the hot leg to the minimum hot leg level currently allowed by operating procedures. At each level, the pump flow will be varied over a wide range. The test data will be analyzed to identify pump operability limits (i.e.,
allowable flow versus hot leg level), as well as the best indicators of SDC performance.
Combustion Engineering is developing instrumentation intended to identify the onset of vortexing in the pump suction line. In cooperation with Combustion Engineering's efforts, we anticipate that this equipment will be included in the instrumentation used in our planned SDC testing at Unit 2. _
ENCLOSURE 1 CALVERT CLIFFS NUCLEAR POWER PLANT FOLLOW-UP RESPONSE TO GENERIC LETTER 88-17 Once the principal SDC performance ' parameters are identified, we will complete our evaluation of the existing instrumentation, and modify existing instrumentation or add new instrumentation as required.. The appropriate alarms-and alarm setpoints for monitoring of SDC system performance will be determined ' based on the testing described above. If necessary, new alarms will be added and setpoints for the existing alarms will be changed.
Following our testing and evaluation of the data, we will provide the NRC. with an additional response which identifies the instruments to be used for ' contint.ously monitoring the SDC system, as well as the associated alarms and alarm setpoints. Any modifications will be completed no later than the end of the 1992 (1991) refueling outages for Unit 1 (Unit 2).
. Our present Control Room 'evel indication and alarm is provided by our refueling level' indication pressure transmitter, which senses off the bottom of one hot leg (opposite the hot leg from which SDC takes suction). We will install additional continuous RCS level detection instrumentation which will - provide Control Room indication and alarm functions. They will be independent, electronic-based instruments utilizing an external standpipe mounting configuration '(see Figure 1). The level. Indication added will be as follows:
o A wide-range, continuous level detection system will be installed on the same hot leg tap as the refueling level indicator. The measurement range will be from the bottom of the hot leg to the Reactor Vessel Flange. A gauge glass will be installed in parallel with the electronic level detection system.
o A narrow-range, continuous level detection system will use other existing taps on an RCS hot leg. Due to the location of the upper and lower taps, the narrow-range level indicator will cover approximately 15 inches above and below the hot leg centerline (approximately 30 inches total span). The independent tap. locations, and the improved accuracy obtained in this I
critical area where we operate (and vortexing could occur), more than compensates for the lack of full range indication across the hot leg.
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