ML20211J525
| ML20211J525 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/30/1999 |
| From: | Beasley J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, LCV-0897-E, NUDOCS 9909030172 | |
| Download: ML20211J525 (5) | |
Text
,
J. Barnia Blastey. Jr.. P.E.-
S:uth:rn Nucle:r Vice President Operating Comp:ny,Inc.
Vogtle Project 40 invemess Center Parkway P0. Box 1295 Birmingham. Alabama 35201
,~
Tel 205 S92.7110 Fax 2059920403 SOUTHERN
' August 30, 1999 COMPANY Energy to Serve Your World" Docket Nos.:
50-424 LCV 0897-E 50-425 U.S. Nuclear Regulatory Commission ATrN: Document Control Desk Washington, D.C. 20555-001 Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT ADDITIONAL INFORMATION CONCERNING GL 96-06, ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENT CONDITIONS By letters dated January 27,1997 and October 28,1998, Southern Nuclear Operating Company (SNC) provided information necessary to address the concerns identified in GL 96-06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions.
Additional information was requested on July 27,1999 relating to the containment penetrations that were identified as being susceptible to thermally-induced pressurization. The requested infonnation is provided as Attachment 1.
Please contact this office if you have any questions.
Sincerely, i
. B. Beasle, Jr JBB/BIIW Attachment ec:
Southern Nuclear Operatine Company Mr. J. T. Gasser Mr. M. Sheibani SNC Document Management U. S. Nuclear Reculatory Commission Mr. L. A. Reyes, Regional Administrator Mr. Ramin R. Assa, Vogtle Project Manager, NRR Mr. J. Zeiler, Senior Resident Inspector, Vogtle 7,'}j iil 7 C 9909030172 990030 ADOCK0500g44 y
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ATTACHMENT 1 Response to GL 96-06 RAI The response to Generic Letter 96-06 identified pipe lines associated with 15 penetrations in each unit that are potentially susceptible to thermally-induced overpressurization.
Procedures were implemented to ensure that isolation of the Reactor Coolant System Hot-Leg Sample Line or the Post-Accident Sampling System Return to Containment is performed only while the fluid is hot during normal operation. For the remaining 13 lines in each unit, interim operability assessments were performed based on an evaluation of isolation valve leakage or an evaluation of the maximum inelastic strain in the pipe. The NRC reviewed the response and requires additional information in order to complete its assessment of the operability of the affected systems and corrective actions.
1.
NRC Request Provide the following information for those penetrations found acceptable based on leakage through isolation valves:
a.
Describe the applicable design criteria for the piping and valves. Include the required load combinations; b.
Provide a drawing of the valve. Provide the pressure at which the valve was determined to lin offits seat or leak and describe the method used to estimate this pressure. Discuss any sources of uncertainty associated with the estimated lift off or leakage pressure; and c.
Provide the maximum-calculated stress in the piping based on the estimated lid offor leakage pressure.
SNC Response Four penetrations were determined to be acceptable based on leakage through
' isolation valves. The information requested for each penetration is described as fcilows.
Accumulator Sample Lines; Penetrations 72A,72B,73A,73B The design conditions for these sections of piping are 700 psig and 300 degrees Fahrenheit. The ANSI pressure temperature rating is 600 pounds.
The required load combinations for the piping and valves are Pressure +
Dead Weight + SSE.
The inboard containment isolation valves (tag numbers 1/2HV-10950,
-10951, -10952, -10953) are 3/4" normally closed solenoid-operated globe valves with flow over the seat. A drawing of these valves is provided as
A Figure 1. The outboard containment isolation valves (tag numbers 1/2-1204-c U6-159, -160, -161, -162) are 3/4" normally closed manually operated globe valves.
The normal pressure and temperature for these sections of piping lines is 650 psig and 120 degrees Fahrenheit. An increase in fluid pressure between these valves would act against the underside of the inboard valve's seat. The maximum differential pressure required to lift the valve disc is 121 psid based on a calculation performed by the valve manufacturer. This is based on the worst case condition where spring forces are maximum and the seat diameter is minimum. There are no additional sources of uncertainty associated with this calculated pressure. Opening of the valve would tend to cause a small amount ofleakage, which would relieve the expansion of the fluid into the corresponding accumulator (tag numbers 1/2-1204-V6-002, -
003,-004,-005). Each of the accumulators are protected from over-pressurization by a relief valve (tag numbers 1/2PSV-8855A, B, C, D), with a setpoint of 700 (i 21) psig, which is vented to the containment atmosphere.
The maximum calculated stress in the piping is based on a pressure of 842 psig (700 + 21 + 121), a pipe outside diameter of 1.050 inches, and a pipe wall thickness of 0.113 inches; along with the required loading combinations.
The calculated stress for these conditions is 33,198 psi. This value is less than the allowable stress of 36,720 psi at AS.ME Section Ill faulted conditions.
2.
NRC Reauest Provide the following additional information for those penetrations found acceptable based on engineering analysis:
a.
Provide the applicable design criteria for the piping and the vai ces. Include the required load combinations; b.
- Provide a drawing of the piping between the isolation valves. Include the lengths and thickness of the piping segments and the type and thickness of the insulation; and c.
Provide the maximum-calculated temperature and pressure for the pipe run.
Describe, in detail, the method used to calculate these pressure and temperature values. This should include a discussion of the heat transfer model used in the analysis and the basis for the heat transfer coefficients used in the analysis.
SNC Response interim operability assessments were performed for 9 penetrations. The acceptability of these assessments was based on evaluation of the maximum inelastic strain in the pipe. However, we did not utilize engineering analysis to demonstrate that a penetration was acceptable on a long-term basis. Further evaluation of these penetrations determined that changes in system operations or modifications were required to alleviate the concerns of potential overpressurization.
Four penetrations were modified to add pressure relief valves between the inboard and outboard containment isolations valves to prevent thermally-induced overpressurization. These four penetrations are described below.
Accumulator Test and Drain Line; Penetration 41 Auxiliary Component Cooling Water (ACCW) Supply; Penetration 28 Normal Containment Sump Pumps Discharge; Penetration 78 e
Reactor Coolant Drain Tank (RCDT) Pump Discharge; Penetration 77 e
Procedures were modified to change system operations for five penetrations to drain any fluid from the piping afler use to ensure that thermally-induced overpressurization is not possible under accident conditions. These five penetrations are described below, Steam Generator Wet Lay-Up Chemical Addition; Penetrations 11 A,12A, e
69A,69B i
Purification Water Supply to Refueling Cavity; Penetration 15 e
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FIGURE I Accumulator Sample Line Inboard Containment Isolation Valve Tag Numbers I/2HV-10950, -10951, -10952, -10953 l'enetrations 72A,72B,73A,73B l
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