ML19351F459
| ML19351F459 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 01/06/1981 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2, TASK-2.K.3, TASK-2.K.3.01, TASK-2.K.3.02, TASK-2.K.3.05, TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM NUDOCS 8101130121 | |
| Download: ML19351F459 (4) | |
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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
Virgil C. Su=cr Nuclear Station Docket No. 50/395 l
TMI Items II.K.2, II.K.3
Dear Mr. Denton:
l South Carolina Electric and Gas Company, acting for itself and agent i.
for South Carolina Public Service Authority, provides forty-five (45) copies i;
of responses to specific items in NUREC-0737. Each item is described below.
I!
(1)
II.K.2.13 - Thermal Mechanical Report -- Effect of high-pressure 1
injection on vessel integrity for small break loss-of-coolant accident with no auxiliary-feedwater To completely address the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during fI recovery from small breaks with an extended loss of all feedwater,-
a Westinghouse Owners Group program will be completed and documented
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to the NRC by January 1,1982. This program will consist of analysis l
for generic (W) PWR plant groupings.
l' Following c.ompletion of this generic program, additional plant specific analyses, if required, will be,provided. A schedule for the plant specific analysis will be determined based on the results j
of the generic analysis.
(2)
II.K.2.17 - Potential for Voiding in the Reactor Coolant Systems During Transient The Westinghouse Owners Group is currently addressing the potential-I for void formation in the reactor coolant system-(RCS) during natural circulation cooldown conditions,- as described in Westinghouse (W) letter NS-TMA-2298 (T. M. Anderson, (W) to P. S. Check, NRC). We believe the resultc of this efforts will fully address the-NRC requirement for__
analysis to determine the potential for' voiding in the RCS'during anti-cipated transients. A report describing the results of this_ effort will be provided to 'the NRC before January 1, '1982.
l 8101180/N g
i
O Mr. Harold R. Denton, January 6, 1981 Page Two (3)
II.K.2.19 - Sequential Auxiliary Feedwater Flow Analysis The transient analysis code, Loftran, and the present small break evaluations analysis code, Wflash, have both undergone benchmarking against plant information or experimental test facilities.
These codes under appropriate conditions have also been compared with each other. The Westinghouse Owners Group will provide on a schedule consistent with requirement of task II.K.2.19 a report addressing the benchmarking of these codes.
(4)
II.K.3.1 - Installation and Testing of Automatic Power-0perated elle alve solation System and II.K.3.2 - Report on Overall Safety Effect of Power-Operated Relief Valve Isolation System The Westinghouse Owners Group is in the process of developing a report (including historical valve failure rate data and documentation of actions taken since the TMI-2 event to decrease the probability of a stuck-open PORV) to address the NRC concerns of item II.K.3.2.
- However, due to the time-consuming process of data gathering, breakdown, and evaluation, this report is scheduled for submittal to the NRC on March 1, 1981. As required by the NiC, this report will be used to support a decision on the necessity of incorporating an automatic PORV isolation system as specified in task action item II.K.3.1.
(5)
II.K.3.5 - Automatic Trip of Reactor Coolant Pump During Loss of Coolant Accident I
The Westinghouse Owners Group resolution of this issue has been to l
perform analyses using the Westinghouse small break evaluation model (Wflash) ta show ample time is available for the operator to trip the reactor coolant pumps following certain size small breaks, see WCAP-9584.
In addition the Owners Group is supporting a best estimate study using the Notrump computer code to demonstrate that tripping the reactor coolant pump at the worst trip time, after a small break will lead to acceptable results.
For both of these analysis efforts, the Westinghouse Owners Group is performing blind post-test predictions of Loft experiment L3-6.
The input data and model to be used with Wflash on Left L3-6 has been submitted to the Staff on 12/1/80 (NS-TMA-2348). The information to be used with Notrump en Loft L3-6 test as stated in letter 00-45 dated 12/3/80.
i
1 Mr. Harold R. Denton January 6, 1981 Page Three (5)
(Con't)
The Loft prediction from both models will be submitted to the Staff on 2/15/81 given that the test is performed on schedule.
l The best estimate study is scheduled for completion by 4/1/81.
Based on these studies, the Westinghouse Owners Group believes' i
that resolution of this issue will be achieved without any design-modifications.
In the event that this is not the case, a schedule will be provided for potential modifications.
?
(6)
II.K.3.30 - Revised Small Break LOCA Methods to Show Compliance with 10CFR50, Appendix K i
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 10CFR Part 50 to be revised, documented, and submitted for NRC approval.
The small break LOCA analysis model currently approved by the NRC for use on Westinghouse designed nuclear ' steam supply systems 1
(including Virgil C. Summer) is conservative and in conformance with Appendix K to 10CFR Part 50.
However, (as documented in letter NS-TMA-2318, dated 9/26/80) Westinghouse believes 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.).
(7)
II.K.3.31 - Plant Specific Calculations to Show Compliance with 10CFR Part 50.46 This item requires that plant specific calculations using the new NRC approved models for small-break LOCAs (refer to item II.K.3.30 above) to show compliance with 10CFR Part 50.46 be submitted for NRC approval.
I As discussed in item II.K.3.30 above (and further documented in letter NS-TMA-2318 dated 9/26/80), Westinghouse is currently working on a schedule to have a new small-break Appendix K model available for use in meeting this requirement. However,.SCE&G's position is that new plant specific analyses utilizing the new and approved Westinghouse model should only be performed and submitted to the NRC if the results of the new model (and subse-quent NRC review and approval) indicate that the present small-break LOCA analyses are not in conformance with 10CFR Part 50.46.
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Mr. Ilarold R. Denton January 6, 1981 Page Four If you have any questions, please let us know.
Very truly yours, ll 6)
T. C. Nichols, Jr.
RBC:TCN:rh cc:
V. C. Summer G. II. Fischer T. C. Nichols, Jr.
E. H. Crews, Jr.
O. W. Dixon, Jr.
W. A. Williams, Jr.
O. S. Bradham D. A. Nauran R. B. Clary A. R. Koon A. A. Smith J. B. Knotts, Jr.
J. L. Skolds B. A. Bursey NPCF/Whitaker File B.
Faas M. Barnisin I
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