ML20058A276
| ML20058A276 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/15/1993 |
| From: | Rehn D DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M87867, TAC-M87868, NUDOCS 9312010026 | |
| Download: ML20058A276 (9) | |
Text
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IDuke Pouer Company D. L Rens
. tatawba Nuclear Generanon Depanment
. Vice11esident '
4&'C ConcordRoad -
(h0?)S313205 Office York,5C2545 (803)S313426 Fax
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DUKEPOWER November 15, 1993 U. S. Nuclear Regulatory Commission g
A'ITN: - Document Control Desk Washington, D. C.
20555
Subject:
Catawba Nuclear Station Docket Nos. 50-413 and 50-414 Response to Request For Additional Information CLA Water Volume And ECCS Subsystem Surveillance Requirements TAC Nos. M87867 AND M87868 i
In a letter dated November 8,1993, the NRC staff requested additional information related to a proposed Technical Specification change submitted by Catawba Nuclear Station on October 5,1993.
This amendment application proposed changes to the cold leg accumulator water volumes, charging and safety injection pump heads and flow rates, and i
residual heat removal pump flow rate. Enclosed is Catawba's response to the questions contained within the request for additional information.
Pursuant to 10 CFR 50.91 (b)(1), the appropriate South Carolina official is being provided a copy of this letter.
Very truly yours,
/Q
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o D. L. Rehn RKS/
J Enclosure 1
9312010026 931115 PDR ADDCK 05000413 g
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1 U. S. Nuclear Regulatory Commission November 15, 1993 t
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xc: Mr. S. D. Ebneter Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Atlanta, GA 30323 Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health &
Environmental Control 2600 Bull Street Columbia, SC 29201 i
American Nuclear Insurers c/o Dottie Sherman, ANI Library The Exchange, Suite 245 270 Farmington Avenue Farmington, CT 06032 M & M Nuclear Consultants 1166 Avenue of the Americas New York, NY-10036-2774 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Mr. R. J. Freudenberger NRC Resident Inspector Catawba Nuclear Station Mr. R. E. Martin Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 14H25 Washington, D.C. 20555 4
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U. S. Nuclear Regulatory Commission Page 3 D. L. Rehn, being duly sworn, states that he is Site Vice-President, Catawba Nuclear Station; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission this revision to the Catawba Nuclear Station Technical Specifications,. Appendix A to License Nos. NPF-35 and NPF-52; and that all statements and matters set forth therein are true and correct to the best of his knowledge.
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L
)gn D. L. Rehn Subscribed and sworn to before me this
/5N day of A/ov.,1993.
b)mm7lber n
blotary Publitf My Commission Expires:
4 A lo v. 2 I, 2 600 i
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DUKE POWER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
APPLICATION FOR AMENDMENTS REGARDING COLD LEG ACCUMULATOR WATER VOLUME AND ECCS SURVEILLANCE REQUIREMENTS Q1.
When and in what documents was Duke Power notified by Westinghouse and Dresser / Pacific Pumps of the changed pump runout limits? Are these findings generic to other users of these pumps? If so how have the other nuclear power plant users been notified?
A.
Duke Power was notified by Westinghouse and Dresser / Pacific in letters DAP-91-074 and DCP-91-074 (Reference 1) dated October 3,1991. The findings identified in Reference I are based upon the size of the pumps and the casting method used tc produce the pump impellers, and are thus generic to other users of these pumps. Duke Power does not have the information necessary to determine how other nuclear power plants were notified on this issue. The NRC should consult with Westinghouse and Dresser / Pacific to determine how other users were notified.
Q2.
Specify the compensatory actions, as mentioned in the October 5,1993 letter, that have been instituted until a Technical Specification (TS) change could be developed.
A.
The centrifugal charging and safety injection systems were evaluated during the injection and sump recirculation phases of a LOCA for flow conditions above the new runout limits identified in Reference 1. For the "as left" centrifugal charging and safety injection system flow balance conditions, runout conditions were not p.--dicted for the safety injection pumps (SIPS), but were predicted for the centrifugal charging pmp (CCPs). As a result, a compensatory action was implemented to isolate reactor coolant pump (RCP) seal injection flow when reactor coolant system (RCS) pressure decreased to 250 psig or below. Isolation of RCP seal injection is only necessary for large break LOCAs when only one centrifugal charging pump is in operation. Isolation of RCP scal injection under these conditions will limit centrifugal charging pump (CCP) flowrates to less than 560 gpm for all modes of operation. Isolation of RCP seal injectian will also allow for more CCP injected flow to be delivered to the cold leg injection lines. In the LOCA analysis, RCP seal injection flow is not modeled.
It was recognized that the long term solution for the immediate CCP runout concern would be to adjust the CLP injected flow balance acceptance criteria such that runout conditions were not predicted for any mode of operation. It was also recognized that the SIP injected flow balance acceptance criteria would need to be revised such that runout conditions would not be predicted if the SIPS were balanced to the upper limits of the flow balance acceptance criteria. However, the modified CCP and SIP flow balance acceptance criteria would provide for slightly less CCP and SIP injected flow than credited in the current LOCA analyses. Therefore, new LOCA analyses were performed to demonstrate the acceptability of slightly less CCP and SIP injected flow. Since the new LOCA results were acceptable, the CCPs and SIPS can be flow balanced within the revised acceptance criteria, and the compensatory action to isolate seal injection can then be removed.
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I Q3.
The change in charging flow, although small, is a decrease while the change in safety injection flow is an increase. Discuss this in view of the problem identified in the vendors notification to the Duke Power Company.
A.
Reference 1 identified new runout limits for the CCPs and SIPS utilized at Catawba Nuclear Station (CNS). The new mnout limits are based on new findings identified by Dresser Pump Division. The CCP runout limit was revised from 565 gpm to 560 gpm.
The SIP runout limit was revised from 660 gpm to 675 gpm. The new runout limits are taken directly from Reference 1. The increase in the SIP runout limit will allow for SIP flowrates above 660 gpm (but less than 675 gpm) during sump recirculation. The proposed CNS Technical Specification changes will limit the CCP and SIP pump flowrates to less than the new runout limits for both the injection and sump recirculation phases of a LOCA.
Q4.
Discuss the generic runout limits and their relationship to the plant data that supported higher limits. Are the limits in the proposed TS consistent with the vendor recommendations or the plant data?
A.
Reference 1 identified revised runout limits for the CCPs as 560 gpm and the SIPS as 675 gpm. The purpose of the proposed CNS Technical Specification changes is to limit pump flowrates to less than these limits for both the injection and sump recirculation phases of a LOCA. Thus, Duke Power intends to implement the vendor recommendations for pump runout. The SIPS were never predicted to have an actual runout problem due to test procedures that limited SIP total flow to 660 gpm during injection line flow balance. Also, plant data from McGuire Nuclear Station (which has the same type of SIP as CNS) indicated that the SIPS could operate safely at approximately 700 gpm. In the Duke Power evaluations, no credit was ever taken for the CCPs to continuously operate above 565 gpm (the old runout limit). The compensatory action to isolate RCP seal injection flow is intended to limit the CCP maximum flowrate to 560 gpm (the new runout limit) or less.
QS.
Discuss the implementation of the charging system compensatory action.- Is it included within procedures and operator training? Does the runout problem exist above RCS pressures of 250 psi? Will the long term solution be implemented duririg.
the Unit 1 EOC-7 outage for both units?
A.
The compensatory action to isolate RCP seal injection flow was implemented under the CNS Compensatory Action Program. Each oncoming shift reviews the compensatory action to ensure all parties involved understand their individual responsibilities and to set up communications as required by the compensatory action. A designated individual (r.)
from cach new shift signs the compensatory action to signify that 1) he/she understands the initial conditions and all instructions /rcsponsibilities for manual operation and 2) he/she can reasonably perform the compensatory action (i.e., strength and accessibility issues).
This particular compensatory action was not included within the CNS ' emergency procedures, nor was it formally included in the operator training program. Isolation of RCP seal injection is not required for RCS pressures above 250 psi because the CCP runout problem does not exist above 250 psi. The long term solution, which is to adjust the CCP and SIP flow balance to prevent runout, will be implemented for Unit I during the EOC-7 outage assuming the proposed Technical Specification changes are approved by the NRC. Thus, the compensatory action to isolate RCP seal injection can be removed for Page 2 of 6
Unit I after the EOC-7 outage. The compensatory action will remain in effect for Unit 2 until a new flow balance can be performed during the next Unit 2 refueling outage. It should be noted that Unit 2 will be in compliance with the proposed Technical Specification changes during the interim.
Q6.
When was the data taken for the development of the pump head curves supporting the proposed TS changes?
A.
The head curves that support the proposed TS changes were selected from the most recent MNS and CNS CCP and SIP head curve data. The weakest CCP and SIP head cunes were selected for developing the LOCA injected flow predictions. Additional degradation was applied to the weakest head curves in developing the injected flow predictions in order to build in conservatism and pump test margin. The strongest CCP and SIP head cunes were selected in evaluating runout conditions for the proposed Technical Specification changes. The test dates for the selected CCP and SIP pump head curves are as follows:
Weakest CCP 12/91 Strongest CCP 5/91 Weakest SIP 9/91 Strongest SIP 7/90 The residual heat removal (RHR) pump head curve that supports the proposed TS changes is based upon the weakest vendor data RHR head curve with additional degradation of approximately 12%. This head curve bounds the weakest RHR pump at MNS or CNS.
Q7.
Specify the "NRC approved LOCA methodology" for the LOCA analysis.
A.
The LOCA reanalyses were perfonned by Westinghouse.
The approved LOCA methodologies are given in WCAP-10266 (Reference 2) and WCAP-10054 (Reference 3).
Q8.
What were the changes in peak cladding temperature for the large and small break LOCAs as a result of the reanalysis?
A.
The current Westinghouse large break and small break peak clad temperatures, as given in Section 15.6.5 of the CNS FSAR (Reference 4), are 1954 and 1440 deg F, respectively.
The large break and small break peak clad temperatures as a result of the reanalysis are 1945 and 1264 deg F, respectively. It should be noted that the LOCA reanalyses incorporated changes other than just ECCS injected flow differences. Rese other changes include, but are not limited to, an increase in the maximum steam generator tube plugging percentage from 10% to 18%, an increase in cold leg accumulator and refueling water storage tank water temperatures, and various Westinghouse LOCA EM improvements.
Herefore, the differences in PCTs cannot be judged solely by the differences in the ECCS injected flow assumptions.
Q9.
Discuss the basis for the change in the residual heat removal flow rate.
A.
As discussed in the answer to the previous question, the LOCA reanalyses incorporated changes other than changes to the ECCS injected flow assumptions. These other changes, which include an increase in the steam generator tube plugging percentage and increases in the accumulator and RWST water temperatures, were expected to be penalties to the final PCTs. The changes in the CCP and SIP injected flow assumptions, which were decreased Page 3 of 6 l
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to limit pump flows to Icss than the revised runout limits provided in Reference 1, were also expected to be PCT penalties. To offset the PCT penalties for the large break analysis, which was closer to the 2200 deg F 10CFR50.46 PCT acceptance criteria, it was decided to take credit for more RHR injected flow. The proposed Technical Specification changes thus reflect this additional RHR injected flow. The latest CNS RHR injected flow test data, which is corrected for uncertainties, indicates that the 3900 gpm proposed TS will be acceptable.
Q10. In regards to the proposed accumulator volume change, specify "the NRC approved methodology."
A.
The required cold leg accumulator volumes are governed by the assumptions utilized in the L.OC.A analyses. The LOCA reanalyses were performed by Westinghouse. The approved LOCA methodologies are given in WCAP-10266 (Reference 2) and WCAP-10054 (Reference 3).
Qll. Discuss the basis for the proposed change in accumulator water volume in terms of instrumentation uncertainties and operational practices. What parameters of the LOCA reanalysis govern the choice of the upper and lower volume limits? Discuss the frequency experienced to date of entry into TS action statements due to the uncertainty limit of plus and minus 20 cubic feet.
A.
The cold leg accumulator level indications are adjusted for instrument uncertainty by increasing or decreasing the allowed indication such that accumulator water volume is maintained within the Technical Specification volume limits. The accumulator high and low level alarms are placed within the allowed level indication band such that the operators are informed of an approaching Technical Specification volume limit. Per Westinghouse, the current BASH-EM supports 130 ft3 andi20 ft3 accumulator water volume uncertainty for MNS and CNS units, respectively, as recorded in the current MNS and CNS Technical Specifications. Since the LOCA reanalyses bound all MNS and CNS units, Westinghouse determined that the accumulator water volume uncertainty for CNS units could be 130 fl3 instead of120 fl3 Per CNS Operations, there have been no recallable entries into TS action statements due to the uncertainty limit of 120 fl3 However, this is primarily because corrective actions would have been undertaken before the accumulator level indication forced entering an action statement. Increasing the accumulator volume tolerance to 130 ft3 will allow for more operating margin, with the potential for fewer required corrective actions. It can also be postulated that the potential for forcez! entries into TS action statements will be less with a larger operating accumulator volume tolerance.
Q12. Duke states that the information provided by Westinghouse and Dresser /PaciCc Pumps indicated that credit would not be taken for an increased pump runout limit
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due to an excess suction pressure, since cavitation is expected to occur on the second i
stage of the pump for flowrates above the initially proposed runout limits. Please provide information on the NPSH limit and the amount of conservatism available on this limit to avoid cavitation.
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A.
During the sump recirculation phase of a LOCA, Duke Power credited the increased suction boost supphed to the CCPs and SIPS by the RHR pumps (approximately 100-150 psig discharge pressure). This increased suction boost was assumed to extend the mnout limits of the CCPs and SIPS to beyond 565 and 660 gpm (old ru'aut limits), respectively, a
during the sump recirculation pluse of a LOCA. Per Reference 1, Westinghouse and
'J Dresser / Pacific Pumps indicated that mnout was limited by cavitation in the second stage of the pumps, and that credit could not be taken for increased suction boost (excess NPSH). CCP and SIP excess NPSH with suction coming from the RWST has never been credited by Duke Power for extending the runout limits beyond 565 and 660 gpm, respectively, during the injection phase of a LOCA. Information on the calculation of NPSH margin can be found in Section 6.3.1 of the CNS FSAR (Reference 4).
Specifically, Table 6-87 lists the required and available NPSH values for the most limitmg conditions.
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REFERENCES 1.
DAP-91-074, DCP-91-074, D. L. Fuller (Westinghouse) to R. C. Futrell (Duke), " Emergency j
Core Cooling System Pump Runout Limit Issues," October 3,1991.
2.
Kabadi, J. N., et al, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Rev. 2, March 1987.
3.
N. Lee, et al, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
4.
Final Safety Analysis Report, Catawba Nuclear Station, October 1,1991.
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