ML18139B665
| ML18139B665 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 12/28/1981 |
| From: | Leasburg R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-2.K.2.17, TASK-2.K.3.05, TASK-2.K.3.25, TASK-2.K.3.30, TASK-TM 702, NUDOCS 8112310339 | |
| Download: ML18139B665 (3) | |
Text
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R.H.LEASBURG VICE PRESIDENT NUCLEAR 0PEllATIONS e
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND~ VIRGINIA 23261 December 28, 1981 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Darrell G. Eisenhut, Director Serial No. 702 NO/DWL:acm Docket Nos. 50-280 50-281 50-338 50-339 Division of Licensing
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Virginia Electric-and-Power Company License Nos.
Gentlemen:
Additional Information on NUREG-0737 Items Being Addressed by Westinghouse Owners Group DPR-32 DPR-37 NFP-4 NFP-7 Several NUREG-0737 items are being addressed generically for Vepco by the Westinghouse Owners Group (WOG).
The status of these items is as follows:
ITEM II.K.2.13 : THERMAL MECHANICAL REPORT-EFFECT OF HIGH PRESSURE INJECTION ON VESSEL INTEGRITY FOR SMALL BREAK LOCA WITH NO AUXILIARY FEEDWATER.
This item requires a detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.
Westinghouse (in support of the Westinghouse Owners Group) is performing an analysis for generic Westinghouse plant groupings to address this issue which will be submitted to the NRC by the end of 1981.
This generic study will be applicable to the Surry and North Anna Units and can be referenced if additional efforts are necessary to completely address NRC concerns (e.g., plant specific analyses).
ITEM II.K.2.17 : POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS Westingthouse (in support of the Westinghouse Owners Group) has performed a study which addresses the potential for void formation in Westinghouse designed nuclear steam supply systems during natural circulation cooldown/depressurization transients.
This study has been submitted to the NRC by the Westinghouse Owners Group (WOG Letter No.
OG-57 dated April 20, 1981) and is applicable to the Surry and North Anna Units.
e VIRGINIA ELECTRIC AND POWER COMPANY TO Harold R. Denton In addition, the Westinghouse Owners Group has developed a natural circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head 2
region during natural circulation cooldown/depressurization transients, and specifies those conditions under which upper head voiding may occur.
These Westinghouse Owners Group generic guidelines have been submitted to the NRC (WOG Letter No. OG-64 dated November 30, 1981).
The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant specific considerations) will be utilized in the implementation of both Surry and North Anna plant specific operat-ing procedures.
ITEM II.K.3.5 : AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT Westinghouse (in support of the Westinghouse Owners Group) has performed an analysis of delayed reactor coolant pump trip during small-break LOCAs.
This analysis is documented in WCAP-9584 (August, 1979).
In addition, Westinghouse (again in support of the Westinghouse Owners Group) has performed test predictions of LOFT Experiments L3-1 and L3-6.
The results of these predictions are documented in WOG Letters OG-49, OG-50, and OG-60 dated March 3, 1981, March 23, 1981 and June 15, 1981, respectively.
Based on:
- 1) the Westinghouse analysis, 2) the favorable prediction of the LOFT Experiment 13-6 results using the Westinghouse analytical model, and 3) Westinghouse simulator data related to operator response time, the Westinghouse position is that automatic reactor coolant pump trip is not necessary since sufficient time is available for manual tripping of the pumps.
Vepco supports this position.
Our understanding of the schedule for final resolution of this issue is:
A)
Once the NRC formally approves the Westinghouse model, a 3-month study period will ensue during which the Westinghouse Owners Group will attempt to demonstrate compliance with NRC acceptance criteria for manual RCP trip.
The NRC acceptance criteria will accompany their formal approval of the Westinghouse models.
B)
If, at the end of the 3-month period, the Westinghouse Owners Group cannot show compliance with the acceptance criteria, the NRC will formally notify utilities that they must submit an automatic RCP trip design.
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VIRGINIA ELECTRIC AND POWER COMPANY TO Harold R. Denton ITEM II.K.3.25 : EFFECT OF LOSS OF AC POWER ON PUMP SEALS This item requires that the consequences of a loss of reactor coolant pump (RCP) seal cooling due to a loss of AC power (defined as loss of offsite power) for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be demonstrated.
During normal operation, seal injection flow from the chemical and volume control system is provided to cool the RCP seals and the component cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals.
In the event of loss of offsite power the RCP motor is deenergized and both of these cooling supplies are terminated; however, the diesel generators are automatically started and seal *injection flow and/or component cooling water to the thermal barrier heat exchanger are automatically restored within seconds.
Either of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of offsite power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ITEM II.K.3.30 : REVISED SMALL-BREAK LOCA METHODS TO SHOW COMPLIANCE WITH 10CFR50, APPENDIX K This item requires that the analysis methods used by NSSS vendors and/or fuel suppliers for small-break LOCA analysis for compliance with Appendix K to 10 CFR Part 50 be revised, documented, and submitted for NRC approval.
Vepco supports the Westinghouse position that the small-break LOCA analysis model currently approved by the NRC is conservative and in conformance with Appendix K to 10 CFR Part 50.
However, as documented in W Letter No. NS-TMA-2318 dated September 26, 1980, Westinghouse b~"iieves that improvement in the realism of small-break calculations is a worthwhile effort and has committed to revise its small-break LOCA analysis model to address NRC concerns (e.g.,
NUREG-0611, NUREG-0623, etc.).
This revised Westinghouse model is currently scheduled for submittal to the NRC by April 1, 1982 as documented in W Letter No. NS-EPR-2524 dated November 25, 1981.
If there are any questions, please contact us at your convenience.
cc:
Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3 Division of Licensing Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing v}J/tl; R.H. Leasburg