ML041670548
| ML041670548 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/03/2004 |
| From: | Salas P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SQN-TS-03-09, TAC MB9513, TAC MB9514 | |
| Download: ML041670548 (4) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 June 3, 2004 SQN-TS-03-09 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) -
UNITS 1 AND 2 -
LICENSE AMENDMENT CHANGE NO.
SQN-TS-03-09 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) (TAC NO. MB9513 AND MB9514)
Reference:
TVA letter to NRC dated June 5, 2003, "Sequoyah Nuclear Plant (SQN) -
Units 1 and 2 -
Proposed License Amendment Request Change No. SQN-TS-03-09 Updated Final Safety Analysis Report (UFSAR)
Failure Modes and Effects Analysis (FMEA) -
Use of Operator Action" TVA submitted TS Change 03-09 to NRC in the referenced letter to propose a change to amend the design and licensing basis to identify that operator action may be necessary to ensure containment design pressure is not exceeded subsequent to a high energy line break such as a loss-of-coolant accident.
Subsequent discussions with NRC indicated that additional information regarding the proposed TS change was needed.
Accordingly, the attached enclosure provides the responses to the additional questions.
There are no commitments contained in this letter.
If you have any questions concerning this change, please contact me at (423) 843-7170 or J.
D. Smith at (423) 843-6672.
Pmeted e receded pay M
U.S. Nuclear Regulatory Commission Page 2 June 3, 2003 I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 3rd day of June, 2004.
ng and Industry Affairs Manager Enclosure cc (Enclosure):
Mr. Michael L. Marshall, Jr., Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop O-8G9A One White Flint North 11555 Rockville Pike-Rockville, Maryland. 20852-2739 Mr. Lawrence E. Nanney, Director Division of Radiological Health Third Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532
ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 DOCKET NOS.
327 AND 328 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
TECHNICAL SPECIFICATION (TS) CHANGE 03-09 RAI Question 1 The June 5, 2003, letter proposes a solution to a high energy line break issue which requires operator action and could result in the simultaneous forced shutdown of the non-accident unit.
Please describe what other solution options were considered or might be considered which would not potentially result in forced shutdown of the non-accident unit and explain why the proposed solution is better.
Response
Actions that were considered to correct this design error were:
- 1) protect the station control and service air (SCSA) system lines inside containment from a pipe whip or jet impingement;
- 2) install motor-operated automatic isolation valves; or 3) rely on operator actions.
As stated-in TVA's letter to NRC dated June 5, 2003, "Sequoyah Nuclear Plant (SQN) -
Units 1 and 2 -
Proposed License Amendment Request Change No. SQN-TS-03 Updated Final Safety Analysis Report (UFSAR) Failure Modes and Effects Analysis (FMEA) -
Use of Operator Action," the risk significance of a high-energy line break (HELB) induced SCSA leak inside containment is estimated for the accident unit based on the increase in large early release frequency (LERF) from this accident scenario.
From the SQN probabilistic safety analysis (PSA), the core damage frequency (CDF) for HELBs is 4.72E-06 per reactor year when combined with the probability of a failure to close the SCSA containment isolation valve; the increase in LERF is 8.OE-09 per reactor year.
The risk significance of a HELB-induced SCSA leak inside containment is estimated for the non-accident unit based on the increase in CDF and LERF due to the higher trip/transient initiation rate as the result of operator actions or errors.
From the SQN PSA, the.conditional core damage probability for a reactor trip is 1.29E-07.
When combined with the probability of a HELB, the increase in CCDF for the non-accident unit is approximately 5.OE-10 per reactor year.
The increase in LERF would be even less.
El-l
A-,
I Because of the low probability of a HELB causing a rupture in a SCSA system line inside containment, coincident with a failure to close the outboard SCSA system containment isolation valve, TVA has chosen to use operator actions to mitigate this event.
The significant cost savings in not performing-,either of the other two actions allows TVA'to focus these resources on equipment reliability and other risk significant issues.
RAI Question 2 Only 40 minutes is allocated in the table on Page El-7 to shutdown the nonaccident unit.
Describe what steps are taken to accomplish this.
What is the final state (TS MODE) of the nonaccident shutdown unit at the end of the 40-minute time interval? Would the SCSA system need to be restored to put the nonaccident unit on the residual heat removal system?
Response
Abnormal operating procedure AOP-C.03, Emergency Shutdown, requires operators to initiate a turbine load reduction at 2 to 5 percent per minute, initiate boration at 20-50 gallons per minute, stop secondary plant equipment (condensate system pumps are stopped between 80 percent and 30ipercent power, one main feedwater pump is stopped at 45 percent power). At a reactor power of 20 percent, the turbine is tripped and then the reactor is tripped.
When the unit is tripped, the plant will be in Hot Standby (MODE 3).
The design and licensing basis for Sequoyah is Mode 3.
In this mode, reactor power will be at zero percent with the reactor trip breakers open, keff is less than 0.99, and T.,g is greater than 350 degrees Fahrenheit.
The SCSA system is not required for operation of the residual heat removal system.
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