RS-04-033, Stations, Units 1 and 2 - Response to RAI Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings
| ML040650522 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/03/2004 |
| From: | Ainger K Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-04-033 | |
| Download: ML040650522 (7) | |
Text
ExekrnSM Exelon Generation www.exeloncorp.com Nuclear 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.90 RS-04-033 March 3, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings
References:
(1) Letter from Kenneth A. Ainger (Exelon Generation Company, LLC) to U.S. NRC, 'Request for a License Amendment to Revise the Pressurizer Safety Valves Lift Settings," dated June 27, 2003 (2)
Letter from Kenneth A. Ainger (Exelon Generation Company, LLC) to U.S. NRC, 'Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings," dated January 29, 2004 In Reference 1, Exelon Generation Company, LLC (EGC) requested NRC approval of a proposed amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. The proposed amendment would revise TS 3.4.10, 'Pressurizer Safety Valves," by changing the existing pressurizer safety valves (PSV) lift setting from "5 2460 psig and
- 2510 psig" to "; 2411 psig and
- 2509 psig" to better reflect the design capabilities of the safety valves while maintaining the appropriate overpressure protection for the reactor coolant system.
During the NRC's review of the proposed change, a number of questions were raised regarding the analyses supporting the revision of the PSV lift setting and the NRC requested that EGC provide additional information to clarify these issues. This information was provided in Reference 2.
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U. S. Nuclear Regulatory Commission March 3, 2004 Page 2 One of the responses (i.e., the response to Question No. 4) provided in Reference 2 addressed the following NRC request:
"Specify the pressure measurement uncertainties associated with the high pressure reactor trip and the PSV, and confirm that they are appropriately considered in the error analysis such that a reactor trip will occurprior to PSV actuation.
Our evaluation of this issue identified that the probability of having a PSV lift (i.e., with the new setpoint of 2460 psig) before achieving a pressurizer pressure - high reactor trip signal (i.e., with a setpoint of 2385 psig) is less than 1 % for any given pressure. Based on this information, the NRC requested that EGC evaluate this potential event to ensure that all accident analyses criteria remain satisfied. Our evaluation of this issue, presented in Attachment 1 to this letter, has confirmed that all applicable accident analysis acceptance criteria remain satisfied.
Also, in Reference 1, we requested that, once approved, the amendment be implemented upon startup of Byron Station Unit 2 Cycle 12 in Spring 2004, Braidwood Station Unit 1 Cycle 12 in Fall 2004, Byron Station Unit 1 Cycle 14 in Spring 2005, and Braidwood Station Unit 2 Cycle 12 in Spring 2005. Based on the current anticipated approval date for this amendment request, we now request that the amendment be implemented upon startup of Byron Station Unit 2 Cycle 13 in Fall 2005 vice the previously requested date for Byron Station Unit 2. There is no change necessary for any of the other units' previously requested implementation dates.
Should you have any questions related to this matter, please contact J. A. Bauer at (630) 657-2801.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on 3
_K __en taX Kenneth A igI Manager, Licensing :
Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings
ATTACHMENT I Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings Introduction In Reference 1, Exelon Generation Company, LLC (EGC) requested NRC approval of a proposed amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. The proposed amendment would revise TS 3.4.10, "Pressurizer Safety Valves," by changing the existing pressurizer safety valves (PSV) lift setting from "Ž 2460 psig and s 2510 psig" to "Ž 2411 psig and
< 2509 psig" to better reflect the design capabilities of the safety valves while maintaining the appropriate overpressure protection for the reactor coolant system (RCS).
During the NRC's review of the proposed change, a number of questions were raised regarding the analyses supporting the revision of the PSV lift setting and the NRC requested that EGC provide additional information to clarify these issues. This information was provided in Reference 2.
One of the responses (i.e., the response to Question No. 4) provided in Reference 2 addressed the following NRC request:
uSpecify the pressure measurement uncertainties associated with the high pressure reactor trip and the PSV, and confirm that they are appropriately considered in the error analysis such that a reactor trip will occur prior to PSV actuation."
Our evaluation of this issue identified that the probability of having a PSV lift (i.e., with the new setpoint of 2460 psig) before achieving a pressurizer pressure - high reactor trip signal (i.e., with a setpoint of 2385 psig) is less than 1% for any given pressure. Based on this information, the NRC requested that EGC evaluate this potential event to ensure that all accident analysis criteria remain satisfied.
Evaluation Specifically, a concern was expressed regarding the proposed change in PSV lift setpoint and tolerance. Assuming the PSVs lift early (i.e., at the low end of the tolerance band) and the high pressurizer pressure reactor trip occurs late (i.e., at the high end of the instrument uncertainty), the PSVs could lift prior to the high pressurizer pressure reactor trip. As a result, the reactor trip on high pressurizer pressure may not occur.
This concern has been evaluated to determine the impact on the appropriate accident analysis acceptance criteria. The results of this evaluation are discussed below.
Of all the analyses documented in the Updated Final Safety Analysis Report (UFSAR),
Chapter 15, "Accident Analysis," only the peak pressure cases for the loss of load/turbine trip (LOL/TT) and rod withdrawal at power (RWAP) analyses resulted in a reactor trip on high pressurizer pressure. In these cases, the PSVs are conservatively modeled to lift at the high end of the tolerance band to maximize the resultant pressure.
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ATTACHMENT I Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings LOlTFT Evaluation For this event, two cases are currently addressed in the analysis of record (AOR). One case is performed to demonstrate meeting the peak pressure acceptance criteria. A separate case is performed to demonstrate meeting the departure from nucleate boiling (DNB) acceptance criterion.
Case 1: Peak Pressure The peak pressure case is performed to demonstrate the primary pressure and secondary pressure meet acceptance criteria. The assumptions used in the AOR are chosen to maximize the resultant pressure. For example, no credit is taken for pressurizer power operated relief valves (PORVs) or pressurizer spray, the PSV lift setpoint is assumed to be at the lift setpoint plus tolerance, the high pressurizer pressure reactor trip setpoint is assumed to be at the reactor trip setpoint plus instrument uncertainty, and a one second delay for loop seal clearance is also assumed.
The following information is obtained from the current AOR for the limiting peak pressure case for the LOL/TT event.
Time (sec)
Event 3.8 pressure corresponding to where the proposed PSV lift setpoint -2%
tolerance would be reached 4.5 pressure corresponding to where the high pressurizer pressure reactor trip setpoint plus uncertainty is reached 5.3 pressure corresponding to where the proposed PSV lift setpoint +2%
tolerance is reached 6.3 steam relief through PSVs begins 6.5 control rod insertion begins based on high pressurizer pressure reactor trip Note that as discussed in Reference 3, a one second delay in steam relief through the PSVs is assumed to account for PSV loop seal clearance. Therefore, steam relief through the PSVs conservatively starts at 6.3 seconds when the lift pressure is reached at 5.3 seconds in the current AOR.
Assuming the PSVs lift at a pressure corresponding to the proposed PSV lift setpoint
-2% tolerance, steam relief through the PSVs would start at 3.8+1 seconds or 4.8 seconds, which is after the setpoint for the high pressurizer pressure reactor trip is reached (i.e., 4.5 seconds). Therefore, the predicted reactor trip signal and timing are not impacted. Assuming the PSVs relieve pressurizer pressure at 4.8 seconds, which is earlier than the time assumed in the current AOR (i.e., 6.3 seconds), the peak pressure results for the proposed change are bounded by the current AOR.
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ATTACHMENT I Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings As described in Byron/Braidwood UFSAR Table 15.0-7, "Plant Systems and Equipment Available for Transients and Accident Conditions," the high pressurizer pressure, overtemperature delta-T (i.e., OTDT), low-low steam generator level, and manual reactor trips are available to mitigate the consequences of a LOLITT event. The AOR predicts high pressurizer pressure as the first reactor trip signal that trips the reactor. If a reactor trip on high pressurizer pressure is not generated, a reactor trip is expected based on OTDT or low-low steam generator level.
Case 2: DNB As discussed in Reference 1, the limiting DNB case for the LOLUTT event has been reanalyzed in support of the proposed change. In accordance with the AOR methodology, the analysis assumptions are chosen to minimize DNBR. For example, full credit is taken for pressurizer PORVs and pressurizer spray, the PSV lift setpoint is assumed to be at the lift setpoint minus tolerance, and the one second delay for loop seal clearance is not assumed. This case minimizes pressure which is conservative for DNBR calculations, as well as pressurizer overfill considerations. The limiting DNB case does not predict a reactor trip from high pressurizer pressure; rather the reactor trip occurs on OTDT. As discussed in Reference 1, the results show that the minimum DNBR value remains well above the safety analysis limit and the pressurizer does not reach a water solid condition. Based on this reanalysis, the concern regarding the PSVs lifting prior to a reactor trip on high pressurizer pressure has no impact on the results of the LOLUTT event limiting DNB case.
RWAP Evaluation For this event, two cases are currently addressed in the AOR. One case is performed to demonstrate meeting the peak pressure acceptance criteria. A separate case is performed to demonstrate meeting the DNB acceptance criterion.
Case 1: Peak Pressure The peak pressure case is performed to demonstrate the primary pressure and secondary pressure meet acceptance criteria. The assumptions used in the AOR are chosen to maximize the resultant pressure. For example, no credit is taken for pressurizer PORVs or pressurizer spray, the PSV lift setpoint is assumed to be at the lift setpoint plus tolerance, the high pressurizer pressure reactor trip setpoint is assumed to be at the reactor trip setpoint plus instrument uncertainty, and a one second delay for loop seal clearance is also assumed.
The following information is obtained from the current AOR for the limiting peak pressure case for the RWAP event.
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ATTACHMENT I Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings Time (sec)
Event 13.5 pressure corresponding to where the proposed PSV lift setpoint -2%
tolerance would be reached 14.3 pressure corresponding to where the high pressurizer pressure reactor trip setpoint plus uncertainty is reached 15.4 pressure corresponding to where the proposed PSV lift setpoint +2%
tolerance is reached 16.3 control rod insertion begins based on high pressurizer pressure 16.4 steam relief through PSVs begins Consistent with the LOLITT analysis, a one second delay in steam relief through the PSVs is assumed to account for PSV loop seal clearance. Therefore, steam relief through the PSVs conservatively starts at 16.4 seconds when the lift pressure is reached at 15.4 seconds in the current AOR.
Assuming the PSVs lift at a pressure corresponding to the proposed PSV lift setpoint
-2% tolerance, steam relief through the PSVs would start at 13.5+1 seconds or 14.5 seconds, which is after the setpoint for the high pressurizer pressure reactor trip is reached. Therefore, the predicted reactor trip signal and timing are not impacted.
Assuming the PSVs relieve pressurizer pressure at 14.5 seconds, which is earlier than the time assumed in the current AOR (i.e., 16.4 seconds), the peak pressure results for the proposed change are bounded by the current AOR.
As described in Byron/Braidwood UFSAR Table 15.0-7, "Plant Systems and Equipment Available for Transients and Accident Conditions," the high pressurizer pressure, OTDT, power range high flux, and manual reactor trips are available to mitigate the consequences of a RWAP event. The AOR predicts high pressurizer pressure as the first reactor trip signal that trips the reactor. If a reactor trip on high pressurizer pressure is not generated, a reactor trip is expected based on OTDT or power range high flux.
Case 2: DNB As discussed in Reference 1, the limiting DNB case for the RWAP event has been reanalyzed in support of the proposed change. In accordance with the AOR methodology, the analysis assumptions are chosen to minimize DNBR. For example, full credit is taken for pressurizer PORVs and pressurizer spray, the PSV lift setpoint is assumed to be at the lift setpoint minus tolerance, and the one second delay for loop seal clearance is not assumed. This case minimizes pressure which is conservative for DNBR calculations, as well as pressurizer overfill considerations. The limiting DNB case does not predict a reactor trip from high pressurizer pressure; rather, the reactor trip occurs on high neutron flux. The results show that the minimum DNBR value remains well above the safety analysis limit and the pressurizer does not reach a water solid condition. Based on this reanalysis, the concern regarding the PSVs lifting prior to a reactor trip on high pressurizer pressure has no impact on the results of the RWAP event limiting DNB case.
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ATTACHMENT I Response to a Request for Additional Information (RAI) Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings The current AOR also analyzed a range of reactivity insertion rates for the RWAP event.
All these cases were examined. The cases producing the largest pressurizer water volume have pressurizer pressure below the proposed PSV setpoint -2% tolerance; therefore, these cases are not impacted by the proposed change. Consequently, the current AOR remains valid and the pressurizer does not reach a water solid condition.
Conclusion The impact from a PSV lifting prior to reaching the high pressurizer pressure reactor trip setpoint has been evaluated for the peak RCS pressure events (i.e., LOLJTT and RWAP). Based on the re-analyses performed to support the proposed PSV setpoint and tolerance change discussed in Reference 1, and evaluation of the current AOR for the LOLITT peak pressure case and the RWAP peak pressure case, it is concluded that all acceptance criteria for these events continue to be met.
References
- 1. Letter from Kenneth A. Ainger (Exelon Generation Company, LLC) to U.S. NRC,
'Request for a License Amendment to Revise the Pressurizer Safety Valves Lift Settings," dated June 27, 2003
- 2. Letter from Kenneth A. Ainger (Exelon Generation Company, LLC) to U.S. NRC, "Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings," dated January 29, 2004
- 3. WCAP 12910, Pressurizer Safety Valve Set Pressure Shift Westinghouse Owners Group Project, MUHP2351/2352," Revision 1-A, May 1993 5