ML041340567
| ML041340567 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/12/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-96-006 | |
| Download: ML041340567 (11) | |
Text
May 12, 2004 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:
In the Matter of )
Docket No. 50-259 Tennessee Valley Authority )
BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 - GENERIC LETTER 96-06, ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENT CONDITIONS This letter responds to NRC Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions," for BFN Unit 1. Generic Letter 96-06 Supplement 1 provided additional guidance but did not require a response.
On September 30, 1996, NRC issued Generic Letter 96-06, which requested licensees address the susceptibility for waterhammer and two-phase flow in the containment air cooler system and over pressurization of piping that penetrates containment. TVA responded for BFN Units 2 and 3 in References 1 through 5. In Reference 6, NRC closed Generic Letter 96-06 for Units 2 and 3.
As detailed in Enclosure 1, TVA has evaluated the Unit 1 containment air cooler cooling water systems to determine if they are susceptible to either water hammer or two-phase flow conditions during postulated accident conditions. TVA has also evaluated piping systems that penetrate containment to determine if they are susceptible to thermal expansion of fluid such that overpressurization of piping may occur. The following conclusions for BFN Unit 1 are consistent with the previous evaluations for Units 2 and 3:
The liquid in the containment cooler coils will not boil during a design basis steam line break or a Loss of Coolant Accident (LOCA) concurrent with a Loss of Offsite Power. Thus, there are no concerns with water hammer or two-phase flow.
U.S. Nuclear Regulatory Commission Page 2 May 12, 2004 The Drywell Floor and Equipment Drains system is acceptable based on leakage through valves which will avoid thermally induced pressure increases above the rated design pressure of the system. However, TVA will modify the system to provide a designed method of overpressure protection.
The Demineralized Water system has the potential to be affected by overpressurization during a postulated LOCA if the piping is completely filled with water and isolated. In response, TVA will implement procedure changes to assure the system is sufficiently drained following use and is open to containment during power operation.
Other water filled systems penetrating primary containment are either not susceptible or can safely accommodate thermally induced overpressurization.
As discussed in NRC Manual Chapter 2509, Browns Ferry Unit 1 Restart Project Inspection Program, since the Generic Letter 96-06 programmatic aspects were verified by NRC inspections for the restarts of Browns Ferry Units 2 and 3, the program does not have to be re-reviewed or re-verified for Unit 1. NRC inspection requirements for the Unit 1 Generic Letter 96-06 program should focus on its implementation.
A detailed discussion of the information requested in Generic Letter 96-06 is contained in Enclosure 1. Enclosure 2 provides a summary of the new commitments contained in this letter and a schedule for their completion.
If you have any questions about this submittal, please contact me at (256) 729-2636.
I declare under penalty of perjury that the foregoing is true and correct. Executed on May 12, 2004.
Sincerely, Original signed by:
T. E. Abney Manager of Licensing and Industry Affairs
U.S. Nuclear Regulatory Commission Page 3 May 12, 2004
References:
- 1.
TVA letter, R. R. Baron to NRC, Browns Ferry Nuclear Plant (BFN)(TAC Nos M96784, M96785, M96786), Sequoyah Nuclear Plant(SQN), and Watts Bar Nuclear Plants (WBN) Unit 1 -
Response to NRC Generic Letter (Generic Letter) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions, Dated September 30, 1996" October 30, 1996.
- 2.
TVA letter, Pedro Salas to NRC, Browns Ferry Nuclear Plant (BFN)(TAC Nos M96784, M96785, M96786), Sequoyah Nuclear Plant(SQN), and Watts Bar Nuclear Plants (WBN) Unit 1 -
Response to NRC Generic Letter (Generic Letter) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions, Dated September 30, 1996" January 28, 1997.
- 3.
TVA letter, T. E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Revision 1, Response to NRC Generic Letter (Generic Letter) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions, October 23, 1997.
- 4.
TVA letter, C. M. Crane to NRC, Browns Ferry Nuclear Plant (BFN) - Request for Additional Information Regarding (RAI)
Response to NRC Generic Letter (Generic Letter) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (TAC Nos. M96784, M96785, M96786), April 15, 1998.
- 5.
TVA letter, T. E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Request for Additional Information Regarding (RAI)
Response and Revised Response to NRC Generic Letter (Generic Letter) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (TAC Nos. M96784, M96785, M96786), August 12, 1998.
- 6.
NRC letter, W. O. Long to TVA, Browns Ferry Units 1,2 & 3, Completion of Licensing Action for Generic Letter 96-06 Concerning WaterHammer, Two-Phase Flow, and Expansion of Entrapped Water in Piping (TAC Nos. M 93436, M93437 and M93438), February 15, 2000.
cc: See Page 4
U.S. Nuclear Regulatory Commission Page 4 May 12, 2004 (Via NRC Electronic Distribution)
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
U.S. Nuclear Regulatory Commission Page 5 May 12, 2004 SMK:BAB cc: M. J. Burzynski, BR 4X-C R. G. Jones, NAB 1A-BFN J. R. Rupert, NAB 1A-BFN K. W. Singer, LP 6A-C M. D. Skaggs, POB 2C-BFN J. Valente, NAB 1E-BFN EDMS S:lic/submit/subs/GL 96 06.doc
E1-1 ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWN FERRY NUCLEAR (BFN) UNIT 1 RESPONSE TO NRC GENERIC LETTER 96-06 ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS This enclosure provides the response to Generic Letter 96-06 for BFN Unit 1.
NRC REQUESTED INFORMATION Licensees were requested to submit a written summary report stating actions taken in response to the requested actions noted below, conclusions that were reached relative to susceptibility for water hammer and two-phase flow in containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued operability of affected systems and components as applicable, and corrective actions that were implemented or planned to be implemented. If systems were found to be susceptible to the conditions that were discussed in the Generic Letter, the affected systems should be identified and the specific circumstances described.
Requested Action(s)
Addressees were requested to determine:
(1)
If containment air cooler cooling water systems are susceptible to either water hammer or two-phase flow conditions during postulated accident conditions; or (2) If piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
In addition to the individual addressees postulated accident conditions, these items should be reviewed with respect to the scenarios referenced in the Generic Letter.
If systems were found to be susceptible to the conditions discussed in the Generic Letter, addressees are expected to assess the operability of the affected systems and take corrective action as appropriate.
E1-2 TVA RESPONSE EVALUATION
SUMMARY
TVA performed an evaluation of each water filled system penetrating primary containment. Each penetration was reviewed for the potential to compromise the containment integrity from thermal expansion of water because of elevated containment temperatures in isolated sections of piping during and following a Loss of Coolant Accident (LOCA). The evaluation determined the maximum pressure that an isolated section of piping would encounter due to elevated temperatures during and following a LOCA would not exceed required structural limits for the affected systems. Thus, containment pressure boundary integrity would be maintained.
EVALUATION DETAILS Sampling System:
The Sampling System was previously analyzed for post-LOCA overpressure utilizing ASME Section III Appendix F methodology.
This system has piping inside primary containment and is normally isolated during power operation. The concern is overpressure of the containment penetration between the inboard and outboard isolation valves. The containment penetration and isolation valves are pneumatic, failed-closed globe valves. These valves are oriented by design such that the above seat is toward the Reactor Pressure Vessel (RPV). Therefore, high pressure between the inboard and outboard valves will unseat the inboard valve and relieve back to the RPV. The containment penetration, affected piping and isolation valves were evaluated and determined to satisfy their structural design basis under the code of record for the system, USAS B31.1.0-1967 with addendums, for the pressure required to lift the valve off the seat and relieve back to the vessel. Therefore, the sampling system was determined to be acceptable.
Main Steam Drain Lines:
The main steam drain lines were originally assumed to be water solid during power operation. These were previously analyzed utilizing ASME Section III Appendix F methodology and found to be acceptable. Operation of the drains was re-reviewed and it was determined that lines would contain only a small amount of water from condensed steam when isolated and would have ample volume for post accident thermal expansion. Therefore, the main steam drain line penetration is not susceptible to thermal overpressurization.
E1-3 Demineralized Water:
This system has the potential to be affected by overpressurization during a postulated LOCA if the piping inside containment is completely filled with water. TVA performed an evaluation which determined that water is normally drained between the outboard isolation valves and a blocking valve during Appendix J testing.
Air retained in the system following the Appendix J testing, prevents the primary containment penetration from being overpressurized during a postulated LOCA. Therefore, because of the as-left configuration, the system will not be affected by the conditions described by the Generic Letter. However, plant procedures will be revised to assure the system is partially drained following use and that piping inside the primary containment is open to the primary containment during power operation.
Standby Liquid Control (SLC) System:
The SLC system does not automatically isolate during a LOCA. The SLC system is open to the reactor vessel. The thermal expansion of water in the primary containment penetration would be relieved back to the vessel through an inboard check valve inside containment. Therefore, the conditions described in the Generic Letter do not pertain to this system.
Drywell Floor and Equipment Drains:
The primary containment isolation valves are both located outside of primary containment. Therefore, the piping between the two valves (and water in the piping) is not subject to the high temperature environment inside containment following a LOCA.
There is a section of piping between the isolation valve outside containment and a flow control valve inside primary containment, which could be exposed to the LOCA temperature environment.
However, the valves inside containment do not provide leak tightness and as such, the maximum pressure is bounded by the design margin, since any thermally induced pressure increase will be bled off through the leaking valves to the sump. Since pressure relief is not a designed function of the valves, a modification will be implemented to provide a designed method of overpressure protection. By calculation it was determined that the total leakage through the check valves must be greater than or equal to 0.5 gallons per minute to mitigate thermally induced pressurization. To ensure this leakage is met, a small orifice will be drilled in each check valve disk.
E1-4 Reactor Water Cleanup System:
The system will be operating normally at the time of the postulated event. Portions of the system piping and penetrations affected by this issue are approximately 500ºF during normal operation. There is no concern for heat gain and pressure increases resulting from temperature increases, as the line will actually cool. Therefore, the system is not susceptible to conditions described in the Generic Letter.
Reactor Building Closed Cooling Water (RBCCW) System:
This system serves the Primary Containment Drywell Coolers. The coolers are used for normal operation to cool primary containment.
The RBCCW system continues to operate post-LOCA as the plant procedures allow the coolers to be used if they are available after an accident. However, accident analyses do not take credit for the coolers.
The RBCCW is a closed loop system with a surge/head tank outside containment. The RBCCW isolation is remote manual from the control room for the discharge line and an in-line check valve for the supply side. RBCCW is not isolated automatically following a LOCA and is not considered part of the containment boundary. As such, there are no primary containment isolation valves that automatically respond to a LOCA or a concurrent Loss of Offsite Power (LOOP). Therefore, the system is not susceptible to the overpressure conditions described in the Generic Letter.
The containment cooler coils were analyzed utilizing the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) 5.0c computer software program. It was determined that the liquid in the containment cooler coils will not boil during a design basis steam line break or LOCA concurrent with a LOOP.
Thus, there will be no water hammer or two-phase flow associated with the restart of the RBCCW pumps.
CONCLUSION The following conclusions for BFN Unit 1 are consistent with the previous evaluations for Units 2 and 3:
The liquid in the containment cooler coils will not boil during a design basis steam line break or a LOCA concurrent with a Loss of Offsite Power. Thus, there are no concerns with water hammer or two-phase flow.
E1-5 The Drywell Floor and Equipment Drains system is acceptable based on leakage through valves which will avoid thermally induced pressure increases above the rated design pressure of the system. However, TVA will modify the system to provide a designed method of overpressure protection.
The Demineralized Water system has the potential to be affected by overpressurization during a postulated LOCA if the piping is completely filled with water and isolated. In response, TVA will implement procedural changes to assure the system is sufficiently drained following use and is open to containment during power operation.
Other water filled system penetrating primary containment are either not susceptible or can safely accommodate thermally induced overpressurization.
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWN FERRY NUCLEAR (BFN) UNIT 1 RESPONSE TO NRC GENERIC LETTER 96-06 ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS COMMITMENT
SUMMARY
- 1.
TVA will modify the Unit 1 Drywell Floor and Equipment Drain Sump discharge lines to provide a designed method of overpressure protection prior to Unit 1 restart.
- 2.
Plant procedures will be revised prior to Unit 1 restart to ensure water is partially drained from portions of the Demineralized Water system inside the drywell and the system left open to the drywell during power operation.