ML20132D633
| ML20132D633 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/25/1985 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8509300242 | |
| Download: ML20132D633 (4) | |
Text
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9 SOUTH CAROLINA ELECTRIC & GAS COMPANY post opriCr to.
CoLUMSIA. south CAROLINA 29218
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Se ptember 25, 1985 VICE PassiotNT NdCLEan opanario%s Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Containment Isolation Valves
Dear Mr. Denton:
South Carolina Electric and Gas Company (SCE6G) submitted a request to revise Table 3.6-1, "Containant Isolat ion Va lves," in a letter from Mr. O. W. Dixon, Jr., to Mr. H. R. Denton dated August 2, 1985.
In discussions with the NRC Staf f, additional information was requested in order to complete the review.
This letter is being provided in response _to that request.
Please note that the questions are answered in two separate attachments to this letter; Attachent A contains discussion on the six seal injection valves in question and Attachment B includes the requested information on the two valves in the Residual Heat Removal (RHR) system. Attachment C contains the marked-up portion of the basis section to the Technical Specifications in support of the information contained in Attachments A and B.
If you should have any questions, please advise.
Yours very truly, b / - [ f t}
- 0. W. Dixon, Jr.
AMM/tdh Attachment cc:
V. C. Summer C. A. Price T. C. Nichols, Jr./0. W. Dixon, Jr.
C. L. Ligon (NSRC)
E. H. Crews, Jr.
K. E. Nodland E. C. Roberts R. A. Stough W. A. Williams, Jr.
G. O. Percival D. A. Nauman C. W. Hehl J. Nelson Grace J. B. Knotts, Jr.
Group Managers H. G. Shealy
- 0. S. Bradham
.NPCF
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File B509300242 850925 PDR ADOCK 05000395 i
PDR gool P
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l ATTACHMENT A Valves XVT-8102 A, B and C-CS XVC-8368 A, B and C-CS Va lve s 8102 A, B, and C are reactor coolant pump seal injection line globe valves located outside of containment. Valves 8368 A, B, and C are seal injection line check valves located inside containment on the same lines as the 8102 valves. These lines are filled with water from the charging system via the charging pucps. These lines sense a pump discharge pressure of 2700 psig.
The seal injection system is normally in service and in operation at all tires, except when the plant is in cold shutdown and reactor coolant pump seal maintenance is required. During an accident condition, for example, a large break LOCA, the seal injection system remains in operation receiving water from the charging pumps and injecting into the Reactor Coolant pump seals. Th is total flow is limited to less than 33 gpm per Technical Specification 3.4.6.2.
During all phases of the accident (injection - cold leg recirculation - hot leg recirculation), the seal injection system continues to inject water into the Reactor Coolant pump seals to ensure the pump seals remain cooled.
If closed during an accident condition, such as for a suspected seal injection line break, the charging pumps will maintain pump discharge pressure against the outside containment s ide of the valve.
In addition, valve leakage through the valve leakof fs would come from the higher pressure side of the valves. This would be the charging pump side (outboard) of the 8201 valves. - The 8368 valves do not have valve leakof fs.
There are no plant procedures which requira closure of the seal injection valves during the first 30 days after onset of the accident.
During a safety injection and subsequent containment isolation, charging pump suction water is provided by the re fueling water storage tank (RWST). Once this supply is exhausted, suction is automatically switched to the reactor building sump. Therefore, s 30 day inventory is always assured. Va lves 8102 A, B, C and 8368 A, B, C are part of the seal water injection portion of the Safety Injection System B Train. Due to normally open cross-tie valves, the Safety Injection System A and B Trains are designed to operate with a single failure of any act ive couponent. Operability of the containment isolation valves is not required to maintain a water seal. Therefore, with the exception of " Loss of Of fsite Power," at least one charging pump is maintaining pressure in both trains (A and B) for all accident scenarios involving a containment isolation signal. Durinp. " Loss of Of fsite Power," the charging pumps will lose power for approximately 16 seconds while the diesel generators are starting. Since the-lines remain filled with water during this period there would not be leakage of containment atmosphere through these lines. All piping and components required to maintain the water seal are ASME safety class 2 or better and seismic Category I.
In addition, charging pump discharge piping is designed for High Energy Line Break (HELB). All components which must operate (charging pumps, valves, etc.) in order to maintain the water seal were designed and purchased as act ive components.
ATTACHMENT B Valves XVG-8701 A and B-RH Valves 8701 A and B provide containment isolation for the Residual Heat Removal (RHR) system pump suctions for the reactor coolant system. They are gate valves located just inside containment.
For these valves, the water seal consists of water filled piping that rises from the inboard s ide of these valves.
For purposes of this analysis, it is assumed that the accident initiating event will break the suction lines just ins ide the missile barrier.
Piping outs ide the missile barrier is not subject to pipe break as a result of the accident initiating event. Since the 8701 valves are ins ide containment, valve leakage from valve leakof fs remains inside the containment. Therefore, the water seal is bounded by the containnent isolation valve and a point level with the bottom of the horizontal run penetrating the missile barrier. The water volume in the piping bounded by these two points is the thirty day inventory. Water leakage from the valve over thirty days must be less than that volume and is to be verified by a leak rate test performed in accordance with station procedures.
These valves are motor operated but have no automatic function to change position under signals such as safety injection or Phase A or B isolation. The valves require operator action to open or close. They also have interlocks to prevent inadvertent operator opening during Modes 1, 2 and 3.
In Mode 4 these valves are opened to initiate RHR.
In all modes of operation, the RHR piping down stream of 8701 A and B is moderate energy. This moderate energy piping ~ is subjected by definition to thru-wall leakage cracks as opposed to guillotine ruptures.
In the event of a thru-wall leakage crack in Mode 4, the RCS system and refueling water storage tank provide inventory for seal source.
Valves 8701 A and B fail as-is and therefore are considered for the Appendix J analysis to be pass ive valves. Therefore, there are no active conponents whose f ailure would af fect the water seal. All piping and components required to maintain the water seal including the vertical riser on the outboard of the valves are ASME safety class 2 or better and seismic Category I.
It is noted that the piping between 8701 and 8702 is designed as high energy piping.
8702 A and B are normally closed valves which also fail in the "as-is" condition. There is not enough energy in the lines between valves 8701 and 8702 to cause a phase A or B isolation. Th e re fore, failure of the water seal piping would not be an accident initiating event.
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3/4.6 CONTAINMENT SYSTEMS BASES
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3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive.
materials from the containment atmosphere will be restricted to those leakage
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paths and associated leak rates assumed in the accident analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
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v.s eck in 3/4.6.1.2 CONTAINMENT LEAKAGE a
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The limitations cn containment leikage rates rensure that the total contain-ment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L or 0.75 L, as applicable, during performance of the periodic test to
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accounfforpossibledegradationofthecontainmentleakagebarriersbetween leakage tests.
s The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.
3/4.6.1.3 REACTOR BUILDING AIR LOCKS The limitations on closure and leak rate for the reactor building air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.
Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
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( u SUMMER - UNIT 1 B 3/4 6-1 9
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