TSTF-10-05, Transmittal of TSTF-523, Revision 0, Generic Letter 2008-01, Managing Gas Accumulation
| ML101800175 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 06/29/2010 |
| From: | Gambrell R, Gregoire D, Raidy T, Schrader K BWR Owners Group, Combustion Engineering, PWR Owners Group, Technical Specifications Task Force, Westinghouse, Babcock & Wilcox Canada, Ltd |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TSTF-10-05, TSTF-523, Rev 0, GL-08-001 | |
| Download: ML101800175 (212) | |
Text
TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY June 29, 2010 TSTF-10-05 PROJ0753 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Transmittal of TSTF-523, Revision 0, "Generic Letter 2008-01, Managing Gas Accumulation" Enclosed for NRC review is TSTF-523, "Generic Letter 2008-01, Managing Gas Accumulation."
TSTF-523 is applicable to all plant types.
Generic Letter (GL) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," addresses the issue of gas accumulation in the emergency core cooling system, decay heat removal system, and containment spray systems to ensure that gas accumulation is maintained less than the amount that challenges Operability of these systems, and that appropriate action is taken when conditions adverse to quality are identified. In the Generic Letter, the Nuclear Regulatory Commission requested that licensees perform evaluations and submit information, including completed actions and schedules for actions to be completed. The GL also states, "the NRC staff plans to use this information during activities that are being planned as a follow-up to this GL and for guidance in the Technical Specifications Task Force program to develop improved TSs."
The TSTF has twice met with the NRC to discuss what Technical Specification (TS) changes might be needed to address the concerns in GL 2008-01. The TSTF and NRC held a public meeting on May 8, 2009 and the TSTF presented our proposed course of action. We met again on November 17, 2009 and the NRC provided the TSTF with a draft document on TS to address gas accumulation issues.
The TSTF has carefully considered the NRC's draft document, which recommends revising or adding Surveillance Requirements and adding a detailed TS Administrative Controls program.
As described in the attached Traveler, the TSTF agrees with the NRC GL explanation that gas management is required to support system operability regardless of whether explicit TS requirements exist. The TSTF believes that the proposed industry approach will enhance plant safety and provides a more comprehensive and flexible approach for resolving this issue.
Addressing the importance of gas management in the TS Bases for the affected systems, 11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation
TSTF 10-05 June 29, 2010 Page 2 describing gas management activities in the UFSAR, and referencing the plant-specific GL 2008-01 responses are sufficient to ensure these technical issues are addressed without imposing additional TS requirements. Furthermore, the TS definition of Operability, Part 9900 of the NRC Inspection Manual, the Maintenance Rule (10 CFR 50.65), the Quality Assurance Program (10 CFR 50, Appendix B), and the NRC's Performance Indicators provide enforcement mechanisms to address gas accumulation, without the need for additional TS requirements.
We suggest a meeting with the NRC early in the review process to discuss the Traveler.
In a separate letter to the NRC's Chief Financial Officer, the TSTF has requested a fee waiver pursuant to the provisions of 10 CFR 170.11 for the review of TSTF-523.
Should you have any questions, please do not hesitate to contact us.
Kenneth J. Schrader (PWROG/W) Donald W. Gregoire (BWROG)
For Thomas W. Raidy (PWROG/CE) Reene' Gambrell (PWROG/B&W)
Enclosure cc: Robert Elliott, Technical Specifications Branch, NRC Barry Miller, Licensing Processes Branch, NRC
WOG-202, Rev. 0 TSTF-523, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Generic Letter 2008-01, Managing Gas Accumulation NUREGs Affected: 1430 1431 1432 1433 1434 Classification 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Improvement NRC Fee Status: Exemption Requested Benefit: Increases Operator Understanding See attached justification.
Revision History OG Revision 0 Revision Status: Active Revision Proposed by: NRC Revision
Description:
Original Issue Owners Group Review Information Date Originated by OG: 28-Mar-10 Owners Group Comments (No Comments)
Owners Group Resolution: Approved Date: 05-Apr-10 TSTF Review Information TSTF Received Date: 10-Jun-10 Date Distributed for Review 29-Jun-10 OG Review Completed: BWOG WOG CEOG BWROG TSTF Comments:
(No Comments)
TSTF Resolution: Approved Date: 29-Jun-10 NRC Review Information NRC Received Date: 29-Jun-10 Affected Technical Specifications LCO 3.4.6 Bases RCS Loops - MODE 4 NUREG(s)- 1430 1431 1432 Only 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 Ref. 3.4.6 Bases RCS Loops - MODE 4 NUREG(s)- 1430 1431 1432 Only LCO 3.4.7 Bases RCS Loops - MODE 5, Loops Filled NUREG(s)- 1430 1431 1432 Only Ref. 3.4.7 Bases RCS Loops - MODE 5, Loops Filled NUREG(s)- 1430 1431 1432 Only LCO 3.4.8 Bases RCS Loops - MODE 5, Loops Not Filled NUREG(s)- 1430 1431 1432 Only Ref. 3.4.8 Bases RCS Loops - MODE 5, Loops Not Filled NUREG(s)- 1430 1431 1432 Only LCO 3.5.2 Bases ECCS - Operating NUREG(s)- 1430 Only Ref. 3.5.2 Bases ECCS - Operating NUREG(s)- 1430 Only Action 3.5.2.A Bases ECCS - Operating NUREG(s)- 1430 Only Action 3.5.2.B Bases ECCS - Operating NUREG(s)- 1430 Only SR 3.5.2.1 Bases ECCS - Operating NUREG(s)- 1430 Only SR 3.5.2.3 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Deleted SR 3.5.2.3 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Deleted SR 3.5.2.4 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.4 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.5 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.5 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.6 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.6 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.7 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.7 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.8 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.7 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.2.8 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.7 SR 3.5.2.9 ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.8 SR 3.5.2.9 Bases ECCS - Operating NUREG(s)- 1430 Only Change
Description:
Renamed SR 3.5.2.8 LCO 3.5.3 Bases ECCS - Shutdown NUREG(s)- 1430 Only SR 3.5.3.1 ECCS - Shutdown NUREG(s)- 1430 Only LCO 3.6.6 Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only Ref. 3.6.6 Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only Action 3.6.6.A Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only SR 3.6.6.4 Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only LCO 3.6.7 Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only Ref. 3.6.7 Bases Containment Spray and Cooling Systems NUREG(s)- 1430 Only LCO 3.9.4 Bases DHR and Coolant Circulation - High Water Level NUREG(s)- 1430 Only Ref. 3.9.4 Bases DHR and Coolant Circulation - High Water Level NUREG(s)- 1430 Only LCO 3.9.5 Bases DHR and Coolant Circulation - Low Water Level NUREG(s)- 1430 Only Ref. 3.9.5 Bases DHR and Coolant Circulation - Low Water Level NUREG(s)- 1430 Only LCO 3.5.2 Bases ECCS - Operating NUREG(s)- 1431 Only Ref. 3.5.2 Bases ECCS - Operating NUREG(s)- 1431 Only Action 3.5.2.A Bases ECCS - Operating NUREG(s)- 1431 Only SR 3.5.2.1 Bases ECCS - Operating NUREG(s)- 1431 Only SR 3.5.2.3 ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Deleted SR 3.5.2.3 Bases ECCS - Operating NUREG(s)- 1431 Only Change
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Deleted SR 3.5.2.4 ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.4 Bases ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.3 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.2.5 ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.5 Bases ECCS - Operating NUREG(s)- 1431 Only Change
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Renamed SR 3.5.2.4 SR 3.5.2.6 ECCS - Operating NUREG(s)- 1431 Only Change
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Renamed SR 3.5.2.5 SR 3.5.2.6 Bases ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.7 ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.7 Bases ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.8 ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.7 SR 3.5.2.8 Bases ECCS - Operating NUREG(s)- 1431 Only Change
Description:
Renamed SR 3.5.2.7 LCO 3.5.3 Bases ECCS - Shutdown NUREG(s)- 1431 Only SR 3.5.3.1 ECCS - Shutdown NUREG(s)- 1431 Only LCO 3.6.6B Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
LCO 3.6.6A Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
LCO 3.6.6C Bases Containment Spray System (Ice Condenser) NUREG(s)- 1431 Only LCO 3.6.6D Bases QS System (Subatmospheric) NUREG(s)- 1431 Only LCO 3.6.6E Bases RS System (Subatmospheric) NUREG(s)- 1431 Only Ref. 3.6.6B Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
Ref. 3.6.6A Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
Ref. 3.6.6C Bases Containment Spray System (Ice Condenser) NUREG(s)- 1431 Only Ref. 3.6.6D Bases QS System (Subatmospheric) NUREG(s)- 1431 Only Ref. 3.6.6E Bases RS System (Subatmospheric) NUREG(s)- 1431 Only SR 3.6.6C.2 Bases Containment Spray System (Ice Condenser) NUREG(s)- 1431 Only 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.6.6D.2 Bases QS System (Subatmospheric) NUREG(s)- 1431 Only SR 3.6.6A.4 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
SR 3.6.6B.4 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
SR 3.6.6B.5 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
SR 3.6.6A.5 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1431 Only Dual)
SR 3.6.6E.5 Bases RS System (Subatmospheric) NUREG(s)- 1431 Only LCO 3.6.7 Bases Spray Additive System (Atmospheric, Subatmospheric, Ice NUREG(s)- 1431 Only Condenser, and Dual)
Ref. 3.6.7 Bases Spray Additive System (Atmospheric, Subatmospheric, Ice NUREG(s)- 1431 Only Condenser, and Dual)
LCO 3.9.5 Bases RHR and Coolant Circulation - High Water Level NUREG(s)- 1431 Only Ref. 3.9.5 Bases RHR and Coolant Circulation - High Water Level NUREG(s)- 1431 Only LCO 3.9.6 Bases RHR and Coolant Circulation - Low Water Level NUREG(s)- 1431 Only Ref. 3.9.6 Bases RHR and Coolant Circulation - Low Water Level NUREG(s)- 1431 Only LCO 3.5.2 Bases ECCS - Operating NUREG(s)- 1432 Only Ref. 3.5.2 Bases ECCS - Operating NUREG(s)- 1432 Only Action 3.5.2.A Bases ECCS - Operating NUREG(s)- 1432 Only Action 3.5.2.B Bases ECCS - Operating NUREG(s)- 1432 Only SR 3.5.2.1 Bases ECCS - Operating NUREG(s)- 1432 Only SR 3.5.2.3 ECCS - Operating NUREG(s)- 1432 Only Change
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Deleted SR 3.5.2.3 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Deleted SR 3.5.2.4 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.4 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.5 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.4 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.2.5 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.6 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.6 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.7 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.7 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.6 SR 3.5.2.8 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.7 SR 3.5.2.8 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.7 SR 3.5.2.9 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.8 SR 3.5.2.9 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.8 SR 3.5.2.10 ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.9 SR 3.5.2.10 Bases ECCS - Operating NUREG(s)- 1432 Only Change
Description:
Renamed SR 3.5.2.9 LCO 3.5.3 Bases ECCS - Shutdown NUREG(s)- 1432 Only SR 3.5.3.1 ECCS - Shutdown NUREG(s)- 1432 Only LCO 3.6.6B Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
LCO 3.6.6A Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
Ref. 3.6.6B Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
Ref. 3.6.6A Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
Action 3.6.6A.A Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
SR 3.6.6B.5 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual) 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.6.6B.6 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
SR 3.6.6A.6 Bases Containment Spray and Cooling Systems (Atmospheric and NUREG(s)- 1432 Only Dual)
LCO 3.6.7 Bases Spray Additive System (Atmospheric and Dual) NUREG(s)- 1432 Only Ref. 3.6.7 Bases Spray Additive System (Atmospheric and Dual) NUREG(s)- 1432 Only LCO 3.9.4 Bases SDC and Coolant Circulation - High Water Level NUREG(s)- 1432 Only Ref. 3.9.4 Bases SDC and Coolant Circulation - High Water Level NUREG(s)- 1432 Only LCO 3.9.5 Bases SDC and Coolant Circulation - Low Water Level NUREG(s)- 1432 Only Ref. 3.9.5 Bases SDC and Coolant Circulation - Low Water Level NUREG(s)- 1432 Only LCO 3.4.8 Bases RHR Shutdown Cooling System - Hot Shutdown NUREG(s)- 1433 Only Ref. 3.4.8 Bases RHR Shutdown Cooling System - Hot Shutdown NUREG(s)- 1433 Only LCO 3.4.9 Bases RHR Shutdown Cooling System - Cold Shutdown NUREG(s)- 1433 Only Ref. 3.4.9 Bases RHR Shutdown Cooling System - Cold Shutdown NUREG(s)- 1433 Only LCO 3.5.1 Bases ECCS - Operating NUREG(s)- 1433 Only Ref. 3.5.1 Bases ECCS - Operating NUREG(s)- 1433 Only Action 3.5.1.A Bases ECCS - Operating NUREG(s)- 1433 Only Action 3.5.1.C Bases ECCS - Operating NUREG(s)- 1433 Only Action 3.5.1.D Bases ECCS - Operating NUREG(s)- 1433 Only Action 3.5.1.E Bases ECCS - Operating NUREG(s)- 1433 Only Action 3.5.1.F Bases ECCS - Operating NUREG(s)- 1433 Only SR 3.5.1.1 ECCS - Operating NUREG(s)- 1433 Only Change
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Deleted SR 3.5.1.1 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Deleted SR 3.5.1.2 ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.1 SR 3.5.1.2 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.1 SR 3.5.1.3 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.2 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.1.3 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.2 SR 3.5.1.4 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.3 SR 3.5.1.4 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.3 SR 3.5.1.5 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.4 SR 3.5.1.5 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.4 SR 3.5.1.6 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.5 SR 3.5.1.6 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.5 SR 3.5.1.7 ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.6 SR 3.5.1.7 Bases ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.6 SR 3.5.1.8 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.7 SR 3.5.1.8 Bases ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.7 SR 3.5.1.9 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.8 SR 3.5.1.9 Bases ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.8 SR 3.5.1.10 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.9 SR 3.5.1.10 Bases ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.9 SR 3.5.1.11 ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.10 SR 3.5.1.11 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.10 SR 3.5.1.12 ECCS - Operating NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.1.11 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.1.12 Bases ECCS - Operating NUREG(s)- 1433 Only Change
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Renamed SR 3.5.1.11 LCO 3.5.2 Bases ECCS - Shutdown NUREG(s)- 1433 Only Ref. 3.5.2 Bases ECCS - Shutdown NUREG(s)- 1433 Only Ref. 3.5.2 Bases RCIC System NUREG(s)- 1433 Only Action 3.5.2.A Bases RCIC System NUREG(s)- 1433 Only SR 3.5.2.3 ECCS - Shutdown NUREG(s)- 1433 Only Change
Description:
Deleted SR 3.5.2.3 Bases ECCS - Shutdown NUREG(s)- 1433 Only Change
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Deleted SR 3.5.2.4 ECCS - Shutdown NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.4 Bases ECCS - Shutdown NUREG(s)- 1433 Only Change
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Renamed SR 3.5.2.3 SR 3.5.2.5 ECCS - Shutdown NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.5 Bases ECCS - Shutdown NUREG(s)- 1433 Only Change
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Renamed SR 3.5.2.4 SR 3.5.2.6 ECCS - Shutdown NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.6 Bases ECCS - Shutdown NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.2.5 LCO 3.5.3 Bases RCIC System NUREG(s)- 1433 Only SR 3.5.3.1 RCIC System NUREG(s)- 1433 Only Change
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Deleted SR 3.5.3.1 Bases RCIC System NUREG(s)- 1433 Only Change
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Deleted SR 3.5.3.2 RCIC System NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.3.1 SR 3.5.3.2 Bases RCIC System NUREG(s)- 1433 Only Change
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Renamed SR 3.5.3.1 SR 3.5.3.3 RCIC System NUREG(s)- 1433 Only Change
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Renamed SR 3.5.3.2 SR 3.5.3.3 Bases RCIC System NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.3.2 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.3.4 RCIC System NUREG(s)- 1433 Only Change
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Renamed SR 3.5.3.3 SR 3.5.3.4 Bases RCIC System NUREG(s)- 1433 Only Change
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Renamed SR 3.5.3.3 SR 3.5.3.5 RCIC System NUREG(s)- 1433 Only Change
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Renamed SR 3.5.3.4 SR 3.5.3.5 Bases RCIC System NUREG(s)- 1433 Only Change
Description:
Renamed SR 3.5.3.4 LCO 3.6.2.3 Bases RHR Suppression Pool Cooling NUREG(s)- 1433 Only Ref. 3.6.2.3 Bases RHR Suppression Pool Cooling NUREG(s)- 1433 Only SR 3.6.2.3.2 Bases RHR Suppression Pool Cooling NUREG(s)- 1433 Only LCO 3.6.2.4 Bases RHR Suppression Pool Spray NUREG(s)- 1433 Only Ref. 3.6.2.4 Bases RHR Suppression Pool Spray NUREG(s)- 1433 Only LCO 3.6.2.4.2 Bases RHR Suppression Pool Spray NUREG(s)- 1433 Only LCO 3.9.8 Bases RHR - High Water Level NUREG(s)- 1433 Only Ref. 3.9.8 Bases RHR - High Water Level NUREG(s)- 1433 Only LCO 3.9.9 Bases RHR - Low Water Level NUREG(s)- 1433 Only Ref. 3.9.9 Bases RHR - Low Water Level NUREG(s)- 1433 Only LCO 3.4.9 Bases RHR Shutdown Cooling System - Hot Shutdown NUREG(s)- 1434 Only Ref. 3.4.9 Bases RHR Shutdown Cooling System - Hot Shutdown NUREG(s)- 1434 Only LCO 3.4.10 Bases RHR Shutdown Cooling System - Cold Shutdown NUREG(s)- 1434 Only Ref. 3.4.10 Bases RHR Shutdown Cooling System - Cold Shutdown NUREG(s)- 1434 Only LCO 3.5.1 Bases ECCS - Operating NUREG(s)- 1434 Only Ref. 3.5.1 Bases ECCS - Operating NUREG(s)- 1434 Only Action 3.5.1.A Bases ECCS - Operating NUREG(s)- 1434 Only Action 3.5.1.B Bases ECCS - Operating NUREG(s)- 1434 Only Action 3.5.1.C Bases ECCS - Operating NUREG(s)- 1434 Only Action 3.5.1.E Bases ECCS - Operating NUREG(s)- 1434 Only Action 3.5.1.F Bases ECCS - Operating NUREG(s)- 1434 Only 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.1.1 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.1.1 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.1.2 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.1 SR 3.5.1.2 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.1 SR 3.5.1.3 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.2 SR 3.5.1.3 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.2 SR 3.5.1.4 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.3 SR 3.5.1.4 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.3 SR 3.5.1.5 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.4 SR 3.5.1.5 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.4 SR 3.5.1.6 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.5 SR 3.5.1.6 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.5 SR 3.5.1.7 ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.6 SR 3.5.1.7 Bases ECCS - Operating NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.1.6 LCO 3.5.2 Bases ECCS - Shutdown NUREG(s)- 1434 Only Ref. 3.5.2 Bases ECCS - Shutdown NUREG(s)- 1434 Only SR 3.5.2.3 ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.2.3 Bases ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.2.4 ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.3 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 SR 3.5.2.4 Bases ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.3 SR 3.5.2.5 ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.5 Bases ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.4 SR 3.5.2.6 ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.5 SR 3.5.2.6 Bases ECCS - Shutdown NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.2.5 LCO 3.5.3 Bases RCIC System NUREG(s)- 1434 Only Ref. 3.5.3 Bases RCIC System NUREG(s)- 1434 Only Action 3.5.3.A Bases RCIC System NUREG(s)- 1434 Only SR 3.5.3.1 RCIC System NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.3.1 Bases RCIC System NUREG(s)- 1434 Only Change
Description:
Deleted SR 3.5.3.2 RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.1 SR 3.5.3.2 Bases RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.1 SR 3.5.3.3 RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.2 SR 3.5.3.3 Bases RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.2 SR 3.5.3.4 RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.3 SR 3.5.3.4 Bases RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.3 SR 3.5.3.5 RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.4 SR 3.5.3.5 Bases RCIC System NUREG(s)- 1434 Only Change
Description:
Renamed SR 3.5.3.4 LCO 3.6.1.7 Bases RHR Containment Spray System NUREG(s)- 1434 Only Ref. 3.6.1.7 Bases RHR Containment Spray System NUREG(s)- 1434 Only 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
WOG-202, Rev. 0 TSTF-523, Rev. 0 LCO 3.6.2.3 Bases RHR Suppression Pool Cooling NUREG(s)- 1434 Only Ref. 3.6.2.3 Bases RHR Suppression Pool Cooling NUREG(s)- 1434 Only SR 3.6.2.3.2 Bases RHR Suppression Pool Cooling NUREG(s)- 1434 Only LCO 3.9.8 Bases RHR - High Water Level NUREG(s)- 1434 Only Ref. 3.9.8 Bases RHR - High Water Level NUREG(s)- 1434 Only LCO 3.9.9 Bases RHR - Low Water Level NUREG(s)- 1434 Only Ref. 3.9.9 Bases RHR - Low Water Level NUREG(s)- 1434 Only 29-Jun-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.
TSTF-523, Rev. 0
- 1. Summary Description Generic Letter (GL) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," (Ref. 1) addresses the issue of gas accumulation in the emergency core cooling system (ECCS), decay heat removal (DHR) (also called Residual Heat Removal (RHR) or Shutdown Cooling (SDC)) system, and containment spray (CS) systems to ensure that gas accumulation is maintained less than the amount that challenges Operability of these systems, and that appropriate action is taken when conditions adverse to quality are identified. The Nuclear Regulatory Commission (NRC) requested that licensees perform evaluations and submit information regarding gas management activities in response to the GL.
The GL states, "the NRC staff plans to use this information during activities that are being planned as a follow-up to this GL and for guidance in the Technical Specifications Task Force program to develop improved TSs."
In response to the Generic Letter, licensees have acknowledged that keeping the subject systems sufficiently full of water is an integral part of the TS definition of Operability. Extensive programmatic and procedural changes are being adopted by licensees in response to GL 2008-01.
As responses to GL 2008-01 are reviewed and closed out by the NRC, licensees will document the need to continue these gas management activities in the future, through incorporation of information into their Updated Final Safety Analysis Reports, including references to the licensee's response to the GL.
The GL stated that the information submitted in response to the GL would be used as guidance by the TSTF to develop the changes for the improved TSs. Based on the ongoing controls that will be in place through this plant-specific licensing basis information and the implementing procedures, combined with the existing regulatory requirements and guidance documents that the NRC detailed within the GL, the TSTF (Technical Specifications Task Force) has developed proposed changes to the Improved Standard Technical Specifications (ISTS) (NUREG-1430 through -1434) (References 2 through 6).
This Traveler utilizes the information to be included in the UFSAR, combined with the TS definition of Operability and additional information to be included in the Bases for the subject systems, to implement actions to manage gas accumulation. This Traveler also includes the proposed information to be included in the UFSAR regarding the management of gas accumulation in the subject systems.
The programmatic and procedural changes implemented by the industry in response to GL 2008-01 address gas accumulation and ensure the Operability of the subject systems in a more complete fashion than the current TS Surveillance Requirements. As discussed below, the existing regulatory requirements are more than adequate to ensure continued adherence to gas management activities that have been put in place. The plant and system-specific nature of the programmatic and procedural changes implemented in response to GL 2008-01, combined with the need for flexibility to incorporate future industry and plant-specific experience, illustrate the appropriateness of including this information in the UFSAR. Changes to the information added to the UFSAR, which will include reference to the licensee-specific responses to GL 2008-01, Page 1
TSTF-523, Rev. 0 would be controlled under 10 CFR 50.59 to address operating experience and the knowledge obtained from ongoing industry programs associated with this issue in a timely manner. When gas is found in one of the subject systems, it is treated as a degraded condition and an Operability determination will be performed. Therefore, this issue is most appropriately addressed in the LCO Bases description of Operability for the system and UFSAR information on periodic system inspections.
Including this information in the UFSAR and referencing the plant-specific GL 2008-01 responses, in conjunction with the TS definition of Operability and a new discussion of the importance of management of gas voids on Operability within the LCO Bases, along with Part 9900 of the NRC Inspection Manual, the Maintenance Rule (10 CFR 50.65), the Quality Assurance Program (10 CFR 50, Appendix B), and the NRC's Performance Indicators are sufficient to ensure these technical issues are addressed without imposing additional TS requirements, and also provide the enforcement mechanism to address gas accumulation.
Specification SR # SR Requirement NUREG Disposition 3.5.2 ECCS- 3.5.2.3 "Verify ECCS piping is full of 1430 (B&W) Remove SR Operating water." 1431 (W) and add Frequency: 31 days 1432 (CE) Operability information to the LCO Bases.
3.5.3 ECCS- 3.5.3.1 References SR 3.5.2.3. 1430 (B&W) Remove SR Shutdown 1431 (W) 3.5.2.3 from 1432(CE) the list in SR 3.5.3.1 and add Operability information to the LCO Bases.
3.5.1 ECCS- 3.5.1.1 "Verify, for each ECCS 1433 (BWR/4) Remove SR Operating injection/spray subsystem, the 1434 (BWR/6) and add piping is filled with water Operability from the pump discharge information to valve to the injection valve." the LCO Frequency: 31 days Bases.
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TSTF-523, Rev. 0 Specification SR # SR Requirement NUREG Disposition 3.5.2 ECCS- 3.5.2.3 "Verify, for each ECCS 1433 (BWR/4) Remove SR Shutdown injection/spray subsystem, the 1434 (BWR/6) and add piping is filled with water Operability from the pump discharge information to valve to the injection valve." the LCO Frequency: 31 days Bases.
3.5.3 RCIC 3.5.3.1 "Verify the RCIC System 1433 (BWR/4) Remove SR System piping is filled with water 1434 (BWR/6) and add from the pump discharge Operability NOTE: The valve to the injection valve." information to Reactor Core Frequency: 31 days the LCO Isolation Bases.
Cooling (RCIC) system is not within the scope of the GL.
However, TS changes for the system are included to maintain consistency within the ISTS.
In addition to the changes to the Specifications listed above, the LCO Bases of the following specifications are modified to add a statement that management of gas voids is important to system Operability and to reference the Gas Management Program in the licensee's UFSAR.
NUREG-1430 (Babcock and Wilcox Plants) 3.4.6, RCS Loops - MODE 4 3.4.7, RCS Loops - MODE 5, Loops Filled 3.4.8, RCS Loops - MODE 5, Loops Not Filled 3.6.6, Containment Spray and Cooling Systems 3.6.7, Spray Additive System 3.9.4, DHR and Coolant Circulation - High Water Level 3.9.5, DHR and Coolant Circulation - Low Water Level NUREG-1431 (Westinghouse Plants) 3.4.6, RCS Loops - MODE 4 3.4.7, RCS Loops - MODE 5, Loops Filled 3.4.8, RCS Loops - MODE 5, Loops Not Filled Page 3
TSTF-523, Rev. 0 3.6.6A, Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6B, Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6C, Containment Spray System (Ice Condenser) 3.6.6D, Quench Spray System (Subatmospheric) 3.6.6E, Recirculation Spray System (Subatmospheric) 3.6.7, Spray Additive System (Atmospheric, Subatmospheric, Ice Condenser, and Dual) 3.9.5, RHR and Coolant Circulation - High Water Level 3.9.6, RHR and Coolant Circulation - Low Water Level NUREG-1432 (Combustion Engineering Plants) 3.4.6, RCS Loops - MODE 4 3.4.7, RCS Loops - MODE 5, Loops Filled 3.4.8, RCS Loops - MODE 5, Loops Not Filled 3.6.6A, Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6B, Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.7, Spray Additive System (Atmospheric and Dual) 3.9.4, SDC and Coolant Circulation - High Water Level 3.9.5, SDC and Coolant Circulation - Low Water Level NUREG-1433 (BWR/4 Plants) 3.4.8, RHR Shutdown Cooling System - Hot Shutdown 3.4.9, RHR Shutdown Cooling System - Cold Shutdown 3.6.2.3, RHR Suppression Pool Cooling 3.6.2.4, RHR Suppression Pool Spray 3.9.8, RHR - High Water Level 3.9.9, RHR - Low Water Level NUREG-1434 (BWR/6 Plants) 3.4.9, RHR Shutdown Cooling System - Hot Shutdown 3.4.10, RHR Shutdown Cooling System - Cold Shutdown 3.6.1.7, RHR Containment Spray System 3.6.2.3, RHR Suppression Pool Cooling 3.9.8, RHR - High Water Level 3.9.9, RHR - Low Water Level In the ISTS NUREGs, subsequent SRs are renumbered to reflect the removed SRs. Note that in license amendment requests to adopt TSTF-523, licensees may choose to mark the removed Surveillances "Not Used" and to not renumber the subsequent SRs. The ISTS NUREG Bases are revised to reflect the change in SR numbering. In NUREG-1431 (W) and NUREG-1432 (CE),
the Bases of TS 3.6.6A and 3.6.6.B, "Containment Spray and Cooling Systems (Atmospheric and Dual)" are revised to correct a reference to an SR in LCO 3.5.2.
It should be noted that NUREG-1432 (Ref. 4) Technical Specification (TS) 3.6.6, "Containment Spray and Cooling Systems" contains SR 3.6.6.4 which states, "Verify the containment spray piping is full of water to the [100] ft level in the containment spray header." This SR was not included in the proposed change because the SR Bases states, "This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containment Page 4
TSTF-523, Rev. 0 analysis." Based on the stated purpose of this SR, it was not included in the changes proposed in this traveler.
The description of requirements to maintain accumulated gas less than the amount that challenges Operability of systems is proposed to be added to the UFSAR. This information is consistent with the licensee-specific responses to GL 2008-01. The information will be added to the UFSAR in accordance with 10 CFR 50.71(e) and will state:
"Gas Management Program The piping systems addressed in the response to NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," have the potential to develop voids and pockets of entrained gases. Maintaining the pump suction and discharge piping sufficiently full of water is necessary to ensure that the system will perform properly and will inject the flow assumed in the safety analyses into the RCS or containment upon demand. This will also prevent damage from pump cavitation or water hammer, and pumping of unacceptable quantities of non-condensable gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following a [Safety Injection (SI)] signal or during shutdown cooling. Consistent with the response to GL 2008-01 dated [DATE] and the NRC's closure of the Generic Letter on [DATE], the procedures implementing the Gas Management Program address the following elements:
- a. Identification of the systems/subsystems for which gas accumulation is managed;
- b. Identification of gas void monitoring locations in suction and discharge piping;
- c. Identification of gas void monitoring frequencies and modifications thereto;
- d. Procedures for documenting and managing activities and data;
- e. Procedures defining corrective actions;
- f. Identification of acceptance criteria; and
- g. Identification of methodology used to determine acceptance criteria."
- 3. Technical Evaluation The TSTF considered the following information to determine the need for and nature of changes to the ISTS to address gas accumulation:
- 1. In the GL, the NRC provided a detailed discussion of the regulatory basis for requiring licensees to ensure that the Operability of the subject systems is not impaired by entrained gas. This discussion referenced the General Design Criteria (GDC) 1, 34, 35, 36, 37, 38, 39, and 40 in Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, and the Quality Assurance (QA) Criteria III, V, XI, XVI, and XVII in 10 CFR 50, Appendix B. (Not all licensees are committed to comply with the GDCs, but may be committed to comply with Page 5
TSTF-523, Rev. 0 similar design requirements stated in their licensing basis.) The NRC also referenced licensee commitments or license requirements to follow other QA documents, such as Regulatory Guide 1.33, "Quality Assurance Requirements (Operation)" (Ref. 7). It was noted that Regulatory Guide 1.33, Appendix A specifically requires instructions for filling and venting the ECCS and DHR systems, as well as for draining and refilling heat exchangers.
- 2. In the GL, the NRC noted that the ISTS and most licensees TS contain SRs to verify that some of the subject systems are full of water. However, the scope of systems to which the SRs apply and the frequency of the verification varies between designs and plants, and some plants do not have similar TS. Furthermore, the current TS requirements to verify that piping is full of water do not correctly reflect the system Operability requirement, since the piping may contain an acceptable amount of entrained gas or voids which do not affect system Operability. As a result, the NRC did not rely on TS compliance as the regulatory basis for the GL.
- 3. The ISTS definition of Operability states, "A system, subsystem, [train/division], component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, [train/division], component, or device to perform its specified safety function(s) are also capable of performing their related support function(s)." (Note that 76 of 104 units have TS consistent with the ISTS definition. The remaining plant's definition of Operability is consistent with this definition in the aspects relevant to this issue.)
There are many aspects of Operability that are required but are not explicitly stated in SRs, such as cooling water, flow balancing, recirculation flow, component integrity, and lubrication. Like entrained gas, these characteristics do not have simple acceptance criteria or fixed verification intervals that lend themselves to Surveillances, but are important characteristics of Operability. Similar periodic verifications of parameters such as motor, pump, or valve stem lubrication are effectively and appropriately controlled in licensee preventive maintenance programs and operating procedure requirements. Should a degraded condition be identified, licensees perform evaluations to determine whether the affected system is Operable.
The system-specific aspects of Operability are described in the LCO Bases. The LCO Bases of the affected Specifications are modified to address the importance of gas management on system Operability.
The NRC has provided detailed inspection guidance to their inspectors for reviewing Operability determinations. This guidance is contained in the NRC Inspection Manual, Part 9900, and was distributed under Regulatory Issue Summary (RIS) 2005-20, Rev. 1, "Revision to NRC Inspection Manual Part 9900 Technical Guidance, 'Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety'."
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TSTF-523, Rev. 0
- 4. In the proceedings for the 1979 Atomic Safety and Licensing Board (ASLB) licensing proceedings for the Trojan Nuclear Power Plant (Portland General Electric Co. Trojan Nuclear Plant, ALAB-531, 9 NRC 263, 273), the NRC's Appeal Board ruled,
"[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."
This position was repeated in the NRC's Final Policy Statement for Technical Specification Improvements for Nuclear Power Plants (58 FR 39182) (the Final Policy Statement), which states, "The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval."
It is not the intent of the NRC that all aspects of Operability be explicitly included in the TS.
As stated in the NRC's Final Policy Statement, " since 1969 there has been a trend towards including in technical specifications not only those requirements derived from the analyses and evaluation included in the plant's safety analysis report but also essentially all other NRC requirements governing the operation of nuclear power plants. In the Commission's view, this has diverted both NRC staff and licensee attention from the more important requirements in these documents to the extent that it has resulted in an adverse but unquantifiable impact on safety."
- 5. There are many instances in which the NRC has identified a potential impact on system Operability, but has not required TS controls to address these impacts. Examples include:
- Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning,"
- Bulletin 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings,"
- Generic Letter 87-12, "Loss of Residual Heat Removal (RHR) While the System (RCS) is Partially Filled,"
- Generic Letter 88-17, "Loss of Decay Heat Removal,"
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TSTF-523, Rev. 0
- Generic Letter 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance,"
- Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment," and
- Generic Letter 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps."
Each affected licensee responded to these issues by revising testing, procedures, preventive maintenance measures, etc., without the adoption of additional TS requirements. Part 9900 of the NRC Inspection Manual, the Maintenance Rule (10 CFR 50.65), the Quality Assurance Program (10 CFR 50, Appendix B), and the NRC's Performance Indicators have been proven sufficient to ensure these technical issues are addressed without imposing additional TS requirements.
The TSTF also considered the following information to determine the appropriate changes to the ISTS NUREGs to address gas accumulation:
- 1. The goal of the TSTF effort was that any change to the TS in response to the GL should be generic and be able to be adopted by licensees without plant-specific variations or the requirement to submit plant-specific evaluations beyond those required to be submitted in response to the GL.
- 2. The evaluations performed by licensees in response to the GL have identified a wide variation in the susceptibility to entrained gas between plants, systems, portions of systems, and operating conditions. As a result, any change to the TS in response to the GL must provide sufficient flexibility for the licensee to adjust the scope (including within systems) and frequency of inspections based on past performance, industry experience, and plant-specific vulnerabilities.
- 3. The acceptance criteria for the volume of entrained gas identified during inspections will vary by plant, system, location, and plant condition. The acceptance criteria will also likely change as additional inspections are performed and mitigating actions and vulnerabilities are identified. Therefore, any change to the TS in response to the GL must provide sufficient flexibility to adjust the acceptance criteria as needed to ensure system Operability.
The TSTF also considered the need to satisfy the 10 CFR 50, Appendix B, Quality Assurance requirements. The following criteria are applicable:
- Criterion III requires measures to ensure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, "Definitions," and as specified in the license, are correctly translated into controlled specifications, drawings, procedures, and instructions.
- Criterion V requires important activities to be prescribed by documented instructions, procedures, or drawings, which must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
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TSTF-523, Rev. 0
- Criterion XI requires a test program to ensure that the subject systems will perform satisfactorily in service. Test results shall be documented and evaluated to ensure that test requirements have been satisfied.
- Criterion XVI requires that measures be established to assure that conditions adverse to quality are promptly identified and corrected.
- Criterion XVII requires that records be maintained to furnish evidence of activities affecting quality.
The programmatic and procedural changes implemented by the industry in response to GL 2008-01 to address gas accumulation in the ECCS, DHR system, and CS system satisfy the applicable criteria of 10 CFR 50, Appendix B.
Based on the requirements of 10 CFR 50 Appendix B, the management of the effect of entrained gas on the Operability of the specified systems would be controlled by the licensee programs and procedures implemented in response to GL 2008-01, in conjunction with the TS requirement that a system is Operable only when it is capable of performing its specified safety functions.
Based on the programmatic and procedural changes implemented by the industry in response to GL 2008-01, the TSTF determined that the current gas accumulation related SRs in the ISTS and plant-specific TS SRs, if any, should be replaced with a gas management program described in the UFSAR. Additionally, as discussed in the responses to GL 2008-01 in Section 3.0 below, the level of detail and complexity associated with the programmatic and procedural changes implemented in response to GL 2008-01 are appropriate to be included in the UFSAR. Changes to the information added to the UFSAR, which will include reference to the licensee-specific responses to GL 2008-01, would be controlled under 10CFR50.59 to address operating experience and the knowledge obtained from ongoing industry programs associated with this issue in a timely manner. When gas is found in one of the subject systems, an Operability determination must be performed. After considering this information, the TSTF concluded that this issue is most appropriately addressed in the UFSAR.
The objective of gas control measures is to limit the volume of gas accumulation to a quantity that does not affect system Operability. An acceptable volume depends on a variety of factors including, but not limited to, the total volume, location, flow rate, type of pump, gas volume fraction at the pump impeller, pressure changes experienced by the system when it is activated, obstacles to flow down stream from the accumulated gas, and the effects of the gas on core cooling. The amount and location of gas are both important in addressing system Operability.
An evaluation to develop and apply criteria is necessary to determine the amount of gas that could impact system Operability.
An SR is inappropriate to address system Operability with respect to the effects of entrained gas on safety systems, as discussed below:
- 1. SRs require that specific verifications be performed at a specified Frequency to ensure that acceptance criteria are met. The evaluation of entrained gas is not consistent with a statement in an SR in that the type of verification will vary (venting, ultrasonic testing, etc.),
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TSTF-523, Rev. 0 the frequency of the verification to be performed and the scope of the systems to be inspected will vary based on past performance and plant-specific vulnerabilities, and the acceptance criteria will vary based on plant, system, and location. For example, some portions of a system may require a monthly verification during particular operating conditions and other portions of the same system may only require post-refueling outage verifications due to low vulnerability and past performance. There would likely be many frequent plant specific exceptions to a proposed SR.
- 2. When the safety analyses assume explicit initial conditions or capabilities of a system such as fluid volume, temperature or pressure, or pump flow, it is appropriate that SRs be provided to verify that these values are within the assumed limits. The current ISTS and TS SRs (References 2 through 6) require verification that the system, or portions of the system, are full of water with no acceptance criteria. In Task Interface Agreement (TIA) 2008-03, dated October 21, 2008 from L. Wert (NRC) to T. Blount (NRC), "Task Interface Agreement -
Emergency Core Cooling System (ECCS) Voiding Relative To Compliance With Surveillance Requirements (SR) 3.5.1.1, 3.5.2.3, and 3.5.3.1," the NRC acknowledged that the intent of the existing TS SRs may be met if the licensee can establish through an Operability determination that there is a reasonable expectation that the system in question will perform its specified safety function."
There are two important conclusions reached from considering this TIA:
- a. There is no acceptance criteria provided in the subject SRs and licensees must determine whether the SR is met based on an Operability determination; and
- b. Regardless of the existence of an SR, licensees must always ensure that systems are Operable when required and must evaluate any degraded condition via an Operability determination. Therefore, an SR provides no additional assurance or benefit.
Additionally, as discussed above, the level of detail and complexity associated with the programmatic and procedural changes implemented in response to GL 2008-01 are appropriate for a UFSAR program.
TS requirements are not needed to ensure inspection and management of the effect of entrained gas on the Operability of the subject systems. As noted above, the NRC did not rely on TS requirements as the basis for consideration of the effect of entrained gas on the Operability of the subject safety systems. Instead, the NRC referenced the 10 CFR 50 General Design Criteria and Quality Assurance Requirements, as well as other requirements. The NRC inspection program provides adequate regulatory oversight of these activities.
Responses to GL 2008-01 Although individual plant responses to GL 2008-01 may vary, there are common elements in the industry guidance for responding to the GL which resulted in the development of plant specific programs and procedures to monitor and mitigate entrained gas in the applicable systems. The following discussions briefly summarize the scope and depth of the programs and procedures already implemented by plants in order to respond to GL 2008-01.
Page 10
TSTF-523, Rev. 0 Licensing Basis Evaluation The Current Licensing Basis was reviewed with respect to gas accumulation for the systems to be evaluated, including periodic venting requirements. The documents reviewed consisted of requirements such as, the TS, TS Bases, UFSAR, Licensee Controlled Documents (e.g.,
Technical Requirements Manual (TRM) and TRM Bases), docketed correspondence, Licensing Commitments, and License Conditions.
In these evaluations, TS requirements such as "Operable flow path" or "Operable train/subsystem" were confirmed to inherently include the requirement that the entrained gas in the flow path or train/subsystem be limited such that the flow path or train/subsystem remains Operable. The proposed addition of an LCO Bases discussion on the importance of management of gas voids on Operability will reinforce this requirement.
Design Evaluation Design Basis Documents were reviewed, including such items as Calculations and Engineering Evaluations and Vendor Technical Manuals, with respect to gas accumulation for the systems to be evaluated.
The results of this review were documented in the response to the GL. The response included a description of any plant specific calculations or analyses that were performed to confirm the acceptability of gas accumulation in the piping of the affected systems, and the acceptance criteria, if applicable.
In addition to the design review, the GL response for the design basis included the following items:
- Develop new applicable gas volume acceptance criteria for each piping segment in each system where gas can accumulate where no acceptance criteria previously existed. This activity included the following:
- 1. Defining the scope of the systems and piping segments involved,
- 2. Pump gas ingestion limits (to ensure pump Operability),
- 3. Pump discharge piping gas limits (to ensure system/piping integrity), and
- 4. RCS Allowable Gas Ingestion limits/analyses (to ensure that the quantities of gas will not prevent the ECCS from performing its core cooling function.
- Review the system P&ID and isometric drawings to identify all system vents and high points.
- Identification of new vent valve locations, modifications to existing vent valves or utilization of existing vent valves that were previously considered to be in inaccessible areas, based on the drawing review.
- Review of the fill and vent activities and procedures for each affected system.
Page 11
TSTF-523, Rev. 0
- Identification of procedure revisions or new procedures resulting from the fill and vent activities and procedure reviews that need to be developed.
- Performance of system confirmation walk downs for the portions of the systems that require venting to ensure that they are sufficiently full of water.
- Identification of new vent valve locations, modifications to existing vent valves, or utilization of existing vent valves that were previously considered to be in inaccessible areas resulting from the confirmatory walk downs.
- Identification of potential gas intrusion mechanisms into each system for each piping segment that is vulnerable to gas intrusion.
Testing Evaluation
- Review the periodic venting or gas accumulation surveillance procedures.
- Identification of procedure revisions, including the surveillance frequency or new procedures that need to be developed as a result of the review.
- Identify how procedures adequately address the manual operation of the RHR/DHR system in its decay heat removal mode of operation. Including how the procedures assure that the RHR/DHR system is sufficiently full of water to perform its decay heat removal safety function (high point venting or ultrasonic testing (UT)) and how pump operation is monitored by plant personnel (including a description of the available instrumentation and alarms).
- Summarize the results of the procedure reviews performed to determine that gas intrusion does not occur as a result of inadvertent draining due to valve manipulations specified in the procedures, system realignments, or incorrect maintenance procedures.
- Describe how gas voids are documented (including the detection method such as venting and measuring or UT and void sizing and post venting checks), dispositioned (including method(s) used such as static or dynamic venting), and trended, if found in any of the subject systems.
- Explain the threshold (acceptance criteria) for entry into the Corrective Action Program (CAP) and how the CAP addresses disposition and trending. For gas voids less than the CAP threshold, if applicable, describe how these gas voids are documented and trended as a means to detect system changes that may be indicative of degradation leading to future gas voiding.
As can be seen from the GL 2008-01 response outline described above, the industry has already implemented sufficient procedural and programmatic controls to monitor and maintain gas accumulation in the affected systems to within acceptable limits.
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TSTF-523, Rev. 0
- 4. Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements/criteria are applicable, as addressed above in Section 3:
- Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TS as part of the license. The Commissions regulatory requirements related to the content for the TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in eight specific categories.
The categories are: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. However, the regulation does not specify the particular requirements to be included in a plants TS.
- Title 10 of the Code of Federal Regulations, Part 50, Paragraph 50.36(c)(3), states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."
- General Design Criteria (GDC) 1, 34, 35, 36, 37, 38, 39, and 40 in Title 10 of the Code of Federal Regulations, Part 50, Appendix A, and the Quality Assurance (QA) Criteria III, V, XI, XVI, and XVII in 10 CFR 50, Appendix B, address aspects of ensuring that systems are not impaired by entrained gas.
The affected SRs are being replaced by a more comprehensive licensee-controlled Gas Management Program contained in the UFSAR. As such, the proposed change does not affect the plant design, hardware, or system operation and will not affect the ability of the plant to perform its design function in mitigating the consequences of a postulated design basis accident.
Therefore, the proposed change does not adversely affect nuclear safety or plant operations.
4.2 No Significant Hazards Consideration Determination The Technical Specifications Task Force (TSTF) has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR that addresses the management of gas and its potential effect on system Operability. The affected SR(s) and gas are not an initiator to any accident previously evaluated. As a result, the Page 13
TSTF-523, Rev. 0 probability of any accident previously evaluated is not significantly increased. The affected systems and components are required to be Operable by the TS and be capable of performing any mitigating function assumed in the safety analysis. Thus, the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR that addresses the management of gas and its potential effect on system Operability. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements that could initiate an accident.
The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR that addresses the management of gas and its potential effect on system Operability. The requirement in the UFSAR to manage gas is more comprehensive than the current TS SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, the TSTF concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Page 14
TSTF-523, Rev. 0 4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.
- 5. Environmental Consideration A review has determined that the proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
- 6. References
- 1. Generic Letter (GL) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," dated January 11, 2008.
- 2. NUREG-1430, "Standard Technical Specifications Babcock and Wilcox Plants," Rev. 3.1, dated December 1, 2005.
- 3. NUREG-1431, "Standard Technical Specifications Westinghouse Plants," Rev. 3.1, dated December 1, 2005.
- 4. NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants," Rev.
3.1, dated December 1, 2005.
- 5. NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4," Rev.
3.1, dated December 1, 2005.
- 6. NUREG-1434, "Standard Technical Specifications General Electric Plants, BWR/6," Rev.
3.1, dated December 1, 2005.
- 7. Regulatory Guide 1.33, "Quality Assurance Requirements (Operation)," Rev. 2, February 1978.
Page 15
TSTF-523, Rev. 0 Attachment 1 Model Application for Adoption of TSTF-523
TSTF-523, Rev. 0
[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
PLANT NAME DOCKET NO. 50-[xxx]
APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-523, "GENERIC LETTER 2008-01, MANAGING GAS ACCUMULATION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].
The proposed amendment would modify TS requirements to address Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," as described in TSTF-523, Revision 0, "Generic Letter 2008-01, Managing Gas Accumulation." [LICENSEE committed to submit this proposed change in
[reference letter].] provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. [Attachment 3 provides revised (clean) TS pages.] Attachment [4] provides existing TS Bases pages marked up to show the proposed changes.
Approval of the proposed amendment is requested by [date]. Once approved, the amendment shall be implemented within [ ] days.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.
Page 1
TSTF-523, Rev. 0
[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.
Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]
If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].
Sincerely,
[Name, Title]
Attachments: 1. Description and Assessment
- 2. Proposed Technical Specification Changes (Mark-Up)
- 3. Revised Technical Specification Pages
- 4. Proposed Technical Specification Bases Changes (Mark-Up)
- 5. Information Copy of Updated Final Safety Analysis Report Gas Management Program cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2
TSTF-523, Rev. 0 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT
1.0 DESCRIPTION
The proposed change revises the Surveillance Requirements which currently require verification that selected portions of systems are full of water. These Surveillance Requirements are deleted and information regarding the importance of management of gas voids on system Operability is added to the Limiting Condition for Operation (LCO) Bases.
The proposed changes add a gas management program to the UFSAR that will be provided to the NRC as required by 10 CFR 50.71(e).
The proposed amendment is consistent with TSTF-523, Revision 0, "Generic Letter 2008-01, Managing Gas Accumulation."
2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation
[LICENSEE] has reviewed the model safety evaluation dated [DATE] as part of the Federal Register Notice of Availability. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-523. [As described in the subsequent paragraphs, ][LICENSEE] has concluded that the justifications presented in the TSTF-523 proposal and the model safety evaluation prepared by the NRC staff are applicable to
[PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the
[PLANT] TS.
2.2 Optional Changes and Variations
[LICENSEE is not proposing any variations or deviations from the TS changes described in the TSTF-523, Revision 0, or the applicable parts of the NRC staffs model safety evaluation dated [DATE].] [LICENSEE is proposing the following variations from the TS changes described in the TSTF-523, Revision 0, or the applicable parts of the NRC staffs model safety evaluation dated [DATE].]
[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-523 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-523 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-523 to the
[PLANT] TS.]
[The [PLANT] TS does not have [all of] the Surveillance Requirements removed from the Standard Technical Specifications on which TSTF-523 was based. This difference does not affect the applicability of TSTF-523 to the [PLANT] TS.]
2.3 Licensee Verifications Page 3
TSTF-523, Rev. 0
[LICENSEE] confirms that information describing a gas management program, consistent with the program described in TSTF-523, has been added into to the Updated Final Safety Analysis Report (UFSAR) and will be incorporated into the UFSAR on the schedule required by 10 CFR 50.71(e). A copy of the gas management program is included for information as .
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination
[LICENSEE] requests adoption of TSTF-523, Rev. 0, "Generic Letter 2008-01, Managing Gas Accumulation," which is an approved change to the standard technical specifications (STS), into the [PLANT NAME, UNIT NOS] technical specifications (TS). The proposed change removes Surveillance Requirements (SRs) that require verifying that systems are full of water and adds information regarding the importance of management of gas voids on system Operability to the Limiting Condition for Operation (LCO) Bases. the proposed change adds a gas management program to the Updated Final Safety Analysis Report (UFSAR).
[LICENSEE] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR that addresses the management of gas and its potential effect on system Operability. The affected SR(s) and gas are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.
The affected systems and components are required to be Operable by the TS and be capable of performing any mitigating function assumed in the safety analysis. Thus, the consequences of any accident previously evaluated are not significantly increased.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR Page 4
TSTF-523, Rev. 0 that addresses the management of gas and its potential effect on system Operability. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements that could initiate an accident. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.
Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change eliminates Surveillance Requirement(s) (SRs) that require verifying that systems are full of water. Information is added to the TS Bases and to the UFSAR that addresses the management of gas and its potential effect on system Operability. The requirement in the UFSAR to manage gas is more comprehensive than the current TS SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.
Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.
Based on the above, [LICENSEE] concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
Page 5
TSTF-523, Rev. 0 - Proposed Technical Specification Changes (Mark-Up)
Page 6
TSTF-523, Rev. 0 - Revised Technical Specification Pages Page 7
TSTF-523, Rev. 0 - Proposed Technical Specification Bases Changes (Mark-Up)
Page 8
TSTF-523, Rev. 0 - Information Copy of Updated Final Safety Analysis Report Gas Management Program Page 9
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 [ Verify the following valves are in the listed position 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ]
with power to the valve operator removed.
Valve Number Position Function
[ ] [ ] [ ]
[ ] [ ] [ ]
[ ] [ ] [ ]
SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.3 [ Verify ECCS piping is full of water. 31 days ]
SR 3.5.2.43 Verify each ECCS pump's developed head at the In accordance test flow point is greater than or equal to the with the Inservice required developed head. Testing Program SR 3.5.2.54 Verify each ECCS automatic valve in the flow path [18] months that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
SR 3.5.2.65 Verify each ECCS pump starts automatically on an [18] months actual or simulated actuation signal.
SR 3.5.2.76 [ Verify the correct settings of stops for the following [18] months ]
HPI stop check valves:
- a. [MUV-2],
- b. [MUV-6], and
- c. [MUV-10].
BWOG STS 3.5.2-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.87 [ Verify the flow controllers for the following LPI [18] months ]
throttle valves operate properly:
- a. [DHV-110] and
- b. [DHV-111].
SR 3.5.2.98 Verify, by visual inspection, each ECCS train [18] months containment sump suction inlet is not restricted by debris and suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.
BWOG STS 3.5.2-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 For all equipment required to be OPERABLE, the In accordance following SRs are applicable: with applicable SRs
[SR 3.5.2.1] SR 3.5.2.65 SR 3.5.2.2 [SR 3.5.2.76]
[SR 3.5.2.3] [SR 3.5.2.87]
SR 3.5.2.43 SR 3.5.2.98 SR 3.5.2.54 BWOG STS 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.6 RCS Loops - MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat to the steam generators (SGs) or decay heat removal (DHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.
In MODE 4, either reactor coolant pumps (RCPs) or DHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RCP or one DHR pump for decay heat removal and transport. The flow provided by one RCP or one DHR pump is adequate for heat removal. The other intent of this LCO is to require that two paths (loops) be available to provide redundancy for heat removal.
APPLICABLE No safety analyses are performed with initial condition in MODE 4.
SAFETY ANALYSES RCS Loops - MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO The purpose of this LCO is to require that two loops, RCS or DHR, be OPERABLE in MODE 4 and one of these loops be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS or DHR System loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. The second loop that is required to be OPERABLE provides redundant paths for heat removal.
Management of gas voids is important to DHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
The Note permits a limited period of operation without RCPs. All RCPs may be removed from operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for the transition to or from the DHR System and otherwise may be de-energized for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. This means that natural circulation has been established using the SGs. The Note prohibits boron dilution with coolant at boron concentrations less than required to assure the SDM of LCO 3.1.1 is maintained when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained at least 10°F below saturation temperature so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.
BWOG STS B 3.4.6-1 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the required DHR or RCS loop in operation to ensure forced flow is providing decay heat removal. Verification includes flow rate, temperature, or pump status monitoring. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval has been shown by operating practice to be sufficient to regularly assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status.
SR 3.4.6.2 Verification that each required pump is OPERABLE ensures that an additional RCS or DHR loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWOG STS B 3.4.6-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO (continued)
Note 2 allows one DHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such testing is safe and possible.
Note 3 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting DHR loops to not be in operation when at least one RCP is in operation. This Note provides for the transition to MODE 4 where an RCP is permitted to be in operation and replaces the RCS circulation function provided by the DHR loops.
An OPERABLE DHR loop is composed of an OPERABLE DHR pump and an OPERABLE DHR heat exchanger.
DHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink when it has an adequate water level and is OPERABLE.
Management of gas voids is important to DHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
APPLICABILITY In MODE 5 with loops filled, forced circulation is provided by this LCO to remove decay heat from the core and to provide proper boron mixing.
One loop of DHR provides sufficient circulation for these purposes.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2,"
LCO 3.4.5, "RCS Loops - MODE 3,"
LCO 3.4.6, "RCS Loops - MODE 4,"
LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled,"
LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level" (MODE 6), and LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level" (MODE 6).
ACTIONS A.1, A.2, B.1, and B.2 If one DHR loop is OPERABLE and any required SG has secondary side water level < [50]% or one required DHR loop inoperable, redundancy for heat removal is lost. Action must be initiated to restore a second DHR loop to OPERABLE status or initiate action to restore the secondary side water level in the SGs, and action must be taken immediately. Either Required Action will restore redundant decay heat removal paths. The BWOG STS B 3.4.7-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.7.3 Verification that each required DHR pump is OPERABLE ensures that redundant paths for heat removal are available. The requirement also ensures that the additional loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation. If the secondary side water level is [50]% in both SGs, this Surveillance is not needed. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWOG STS B 3.4.7-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO (continued)
Note 1 permits the DHR pumps to be removed from operation for 15 minutes when switching from one train to the other. The circumstances for stopping both DHR pumps are to be limited to situations where the outage time is short [and temperature is maintained
[160]°F]. The Note prohibits boron dilution with coolant at boron concentrations less than required to assure the SDM of LCO 3.1.1 is maintained or draining operations when DHR forced flow is stopped.
Note 2 allows one DHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.
An OPERABLE DHR loop is composed of an OPERABLE DHR pump capable of providing forced flow to an OPERABLE DHR heat exchanger.
DHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.
Management of gas voids is important to DHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the DHR System.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2,"
LCO 3.4.5, "RCS Loops - MODE 3,"
LCO 3.4.6, "RCS Loops - MODE 4,"
LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"
LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level" (MODE 6), and LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level" (MODE 6).
ACTIONS A.1 If one required DHR loop is inoperable, redundancy for heat removal is lost. Required Action A.1 is to immediately initiate activities to restore a second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
BWOG STS B 3.4.8-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES REFERENCES 1. Generic Letter 88-17, October 17, 1988.
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWOG STS B 3.4.8-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES APPLICABLE SAFETY ANALYSES (continued)
In the LOCA analyses, HPI and LPI are not credited until 35 seconds after actuation of the ESFAS signal. This is based on a loss of offsite power and the associated time delays in startup and loading of the emergency diesel generator (EDG). Further, LPI flow is not credited until RCS pressure drops below the pump's shutoff head. For a large break LOCA, HPI is not credited at all.
The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that at least one is available, assuming a single failure in the other train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.
In MODES 1, 2, and 3, an ECCS train consists of an HPI subsystem and an LPI subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST upon an ESFAS signal and manually transferring suction to the containment sump.
During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the BWST to the RCS via the HPI and LPI pumps and their respective discharge flow paths to each of the four cold leg injection nozzles and the reactor vessel. In the long term, this flow path may be manually transferred to take its supply from the containment sump and to supply its flow to the RCS via two paths, as described in the Background section.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 3).
The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.
As indicated in the Note, operation in MODE 3 with ECCS trains de-activated pursuant to LCO 3.4.12 is necessary for plants with an LTOP System arming temperature at or near the MODE 3 boundary temperature of [350]°F. LCO 3.4.12 requires that certain components be de-activated at and below the LTOP System arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to restore the systems to OPERABLE status.
BWOG STS B 3.5.2-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES APPLICABILITY In MODES 1, 2, and 3, the ECCS train OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPI pump performance is based on the small break LOCA, which establishes the pump performance curve and is less dependent on power. The HPI pump performance requirements are based on a small break LOCA. MODES 2 and 3 requirements are bounded by the MODE 1 analysis.
In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops -
MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level."
ACTIONS A.1 With one LPI subsystem inoperable, action must be taken to restore it to OPERABLE status within 7 days. In this condition, the remaining OPERABLE ECCS train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure to the remaining LPI subsystem could result in loss of ECCS function. The [7] day Completion Time is reasonable to perform corrective maintenance on the inoperable LPI subsystem. The [7] day Completion Time is based on the findings of the deterministic and probabilistic analysis in Reference 43. Reference 43 concluded that extending the Completion Time to [7] days for an inoperable LPI subsystem proves plant operational flexibility while simultaneously reducing overall plant risk. This is because the risks incurred by having the LPI subsystem unavailable for a longer time at power will be substantially offset by the benefits associated with avoiding unnecessary plant transitions and by reducing risk during plant shutdown operations.
B.1 With one or more trains operable and at least 100% of the injection flow equivalent to a single OPERABLE ECCS train available, components inoperable for reasons other than Condition A must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on NRC recommendations (Ref. 54) that are based on a risk evaluation and is a reasonable time for many repairs.
BWOG STS B 3.5.2-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES ACTIONS (continued)
An ECCS train is inoperable if it is not capable of delivering the design flow to the RCS.
The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.
An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 54) has shown the risk of having one full ECCS train inoperable to be sufficiently low to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
With one or more components inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be immediately entered.
C.1 and C.2 If the inoperable components cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 Condition A is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available.
With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the facility is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
BWOG STS B 3.5.2-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power or by key locking the control in the correct position ensures that the valves cannot change position as the result of an active failure. These valves are of the type described in Reference 65, which can disable the function of both ECCS trains and invalidate the accident analyses. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in view of other administrative controls that will ensure the unlikelihood of a mispositioned valve.
SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under administrative control, and an inoperable valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.2.3 With the exception of systems in operation, the ECCS pumps are normally in a standby, nonoperating mode. As such, the flow path piping has the potential to develop voids and pockets of entrained gases.
Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an ESFAS signal or during shutdown cooling. The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the existence of procedural controls governing system operation.
BWOG STS B 3.5.2-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.43 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code (Ref. 76). This type of testing may be accomplished by measuring the pump's developed head at only one point of the pump's characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant accident analysis. SRs are specified in the Inservice Testing Program of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
SR 3.5.2.5 4and SR 3.5.2.65 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated ESFAS signal and that each ECCS pump starts on receipt of an actual or simulated ESFAS signal.
This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the ESFAS testing, and equipment performance is monitored as part of the Inservice Testing Program.
SR 3.5.2.76 This Surveillance ensures that these valves are in the proper position to prevent the HPI pump from exceeding its runout limit. This 18 month Frequency is based on the same reasons as those stated for SR 3.5.2.5 4 and SR 3.5.2.65.
BWOG STS B 3.5.2-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.87 This Surveillance ensures that the flow controllers for the LPI throttle valves will automatically control the LPI train flow rate in the desired range and prevent LPI pump runout as RCS pressure decreases after a LOCA. The 18 month Frequency is justified by the same reasons as those stated for SR 3.5.2.5 4 and SR 3.5.2.65.
SR 3.5.2.98 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to preserve access to the location, and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation and has been confirmed by operating experience.
REFERENCES 1. 10 CFR 50.46.
- 2. FSAR, Section [6.3].
- 3. FSAR, Section [ ], ["Gas Management Program."]
- 43. BAW-2295-A, Revision 1, Justification for Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray System.
- 54. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
- 65. IE Information Notice 87-01, "RHR Valve Misalignment Causes Degradation of ECCS in PWRs," January 6, 1987.
- 76. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWOG STS B 3.5.2-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications.
In MODE 4, the required ECCS train consists of two separate subsystems: high pressure injection (HPI) and low pressure injection (LPI), each consisting of two redundant, 100% capacity trains.
The ECCS flow paths consist of piping, valves, heat exchangers, and pumps, such that water from the borated water storage tank (BWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.
APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 is applicable to SAFETY these Bases.
ANALYSES Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. Included in these reductions is that certain automatic Engineered Safety Feature Actuation System (ESFAS) actuation is not available. In this MODE sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
Only one ECCS train is required for MODE 4. This requirement dictates that single failures are not considered during this MODE. The ECCS train
- shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to ensure sufficient ECCS flow is available to the core following a DBA.
In MODE 4, an ECCS train consists of an HPI subsystem and an LPI subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST and transferring suction to the containment sump.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report.
During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the BWST to the RCS, via the ECCS pumps and their respective supply headers, to each of the four BWOG STS B 3.5.3-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Containment Spray and Cooling Systems B 3.6.6 BASES APPLICABLE SAFETY ANALYSES (continued)
The modeled Containment Cooling System actuation from the containment analysis is based on a response time associated with exceeding the containment pressure high setpoint to achieve full Containment Cooling System air and safety grade cooling water flow.
The Containment Cooling System total response time of [25] seconds includes signal delay, DG startup (for loss of offsite power), and service water pump startup times (Ref. 3).
The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, a minimum of one containment cooling train and one containment spray train are required to maintain the containment peak pressure and temperature below the design limits. Additionally, one containment spray train is required to remove iodine from the containment atmosphere and maintain concentrations below those assumed in the safety analysis. To ensure that these requirements are met, two containment spray trains and two containment cooling units must be OPERABLE. Therefore, in the event of an accident, the minimum requirements are met, assuming the worst-case single active failure occurs.
Each Containment Spray System typically includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST upon an Engineered Safety Features Actuation System signal and manually transferring suction to the containment sump.
Management of gas voids is important to Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 5).
Each Containment Cooling System typically includes demisters, cooling coils, dampers, an axial flow fan driven by a two speed water cooled electrical motor, instruments, and controls to ensure an OPERABLE flow path.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature, requiring the operation of the containment spray trains and containment cooling trains.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray System and the Containment Cooling System are not required to be OPERABLE in MODES 5 and 6.
BWOG STS B 3.6.6-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems B 3.6.6 BASES ACTIONS A.1 With one containment spray train inoperable, action must be taken to restore it to OPERABLE status within [7] days. In this condition, the remaining OPERABLE containment spray train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure to the remaining containment spray train could result in loss of spray function. The [7] day Completion Time is reasonable to perform corrective maintenance on the inoperable containment spray train. The [7] day Completion Time is based on the findings of the deterministic and probabilistic analysis in Reference 65. Reference 65 concluded that extending the Completion Time to [7] days for an inoperable containment spray train proves plant operational flexibility while simultaneously reducing overall plant risk. This is because the risks incurred by having the containment spray train unavailable for a longer time at power will be substantially offset by the benefits associated with avoiding unnecessary plant transitions and by reducing risk during plant shutdown operations.
B.1 and B.2 If the inoperable containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time to attempt restoration of the containment spray train and is reasonable when considering the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.
C.1 With one of the required containment cooling trains inoperable, the inoperable containment cooling train must be restored to OPERABLE status within 7 days. The components in this degraded condition provide iodine removal capabilities and are capable of providing at least 100% of the heat removal needs after an accident. The 7 day Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System and the low probability of a DBA occurring during this period.
BWOG STS B 3.6.6-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the containment spray flow path provides assurance that the proper flow paths will exist for Containment Spray System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those valves outside containment and capable of potentially being mispositioned are in the correct position.
SR 3.6.6.2 Operating each [required] containment cooling train fan unit for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling trains occurring between surveillances and has been shown to be acceptable through operating experience.
SR 3.6.6.3 Verifying that each [required] containment cooling train provides an essential raw water cooling flow rate of [1780] gpm to each cooling unit provides assurance that the design flow rate assumed in the safety analyses will be achieved (Ref. 1). The Frequency was developed considering the known reliability of the Cooling Water System, the two train redundancy available, and the low probability of a significant degradation of flow occurring between surveillances.
SR 3.6.6.4 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 76). Since the Containment Spray System pumps cannot be tested with flow through the BWOG STS B 3.6.6-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems B 3.6.6 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 2. FSAR, Section [14.1].
- 3. FSAR, Section [6.3].
- 4. FSAR, Section [14.2].
- 5. FSAR, Section [ ], ["Gas Management Program."]
- 65. BAW-2295-A, Revision 1, Justification for Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray Systems.
- 76. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWOG STS B 3.6.6-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Spray Additive System B 3.6.7 BASES APPLICABLE SAFETY ANALYSES (continued)
The DBA response time assumed for the Spray Additive System is the same as for the Containment Spray System and is discussed in the Bases for LCO 3.6.6, "Containment Spray and Cooling Systems."
The DBA analyses assume that one train of the Containment Spray System/Spray Additive System is inoperable and that the entire spray additive tank volume is added to the remaining Containment Spray System flow path.
In the evaluation of the worst-case LOCA, the safety analysis assumed that an alkaline containment spray effectively reduced the airborne iodine.
Each Containment Spray System suction line is equipped with its own gravity feed from the spray additive tank. Therefore, in the event of a single failure within the Spray Additive System (i.e., suction valve failure),
NaOH will still be mixed with the borated water, establishing the alkalinity essential to effective iodine removal.
The Spray Additive System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO The Spray Additive System is necessary to reduce the release of radioactive material to the environment in the event of a DBA. To be considered OPERABLE, the volume and concentration of the spray additive solution must be sufficient to provide NaOH injection into the spray flow until the Containment Spray System suction path is switched from the BWST to the containment sump and to raise the average spray solution pH to a level conducive to iodine removal. The average spray solution pH is between [7.2 and 11.0]. This pH range maximizes the effectiveness of the iodine removal mechanism without introducing conditions that may induce caustic stress corrosion cracking of mechanical system components. In addition, it is essential that valves in the Spray Additive System flow paths are properly positioned and that automatic valves are capable of activating to their correct positions.
Management of gas voids is important to Spray Additive System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment requiring the operation of the Spray Additive System. The Spray Additive System assists in reducing the iodine fission product inventory prior to release to the environment.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these BWOG STS B 3.6.7-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Spray Additive System B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the Containment Spray System, the flow [rate] in the Spray Additive System is verified once per 5 years. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation. Due to the passive nature of the spray additive flow controls, the 5 year Frequency is sufficient to identify component degradation that may affect flow [rate].
References 1. FSAR, Section [6.2].
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWOG STS B 3.6.7-6 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 DHR and Coolant Circulation - High Water Level B 3.9.4 BASES LCO Only one DHR loop is required for decay heat removal in MODE 6, with a water level 23 ft above the top of the reactor vessel flange. Only one DHR loop is required to be OPERABLE because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one DHR loop must be OPERABLE and in operation to provide:
- a. Removal of decay heat,
- b. Mixing of borated coolant to minimize the possibility of criticality, and
- c. Indication of reactor coolant temperature.
An OPERABLE DHR loop includes a DHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to DHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
Additionally, each DHR loop is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation of one subsystem can maintain the reactor coolant temperature as required.
The LCO is modified by a Note that allows the required DHR loop to be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1. Boron concentration reduction with coolant at boron concentrations less than required to assure the RCS boron concentration is maintained is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to DHR isolation valve testing. During this 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.
BWOG STS B 3.9.4-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 DHR and Coolant Circulation - High Water Level B 3.9.4 BASES ACTIONS (continued)
A.3 If DHR loop requirements are not met, actions shall be initiated immediately in order to satisfy DHR loop requirements.
A.4, A.5, A.6.1, and A.6.2 If no DHR is in operation, the following actions must be taken:
- a. The equipment hatch must be closed and secured with [four] bolts,
- b. One door in each air lock must be closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.
With DHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most DHR problems and is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the DHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the DHR System.
REFERENCES 1. FSAR, Section [ ] .
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWOG STS B 3.9.4-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 DHR and Coolant Circulation - Low Water Level B 3.9.5 BASES LCO (continued)
This LCO is modified by two Notes. Note 1 permits the DHR pumps to be removed from operation for 15 minutes when switching from one train to another. The circumstances for stopping both DHR pumps are to be limited to situations when the outage time is short [and the core outlet temperature is maintained > 10 degrees F below saturation temperature].
The Note prohibits boron dilution of draining operations by introduction of coolant into the RCS with boron concentrations less than required to meet the minimum boron concentration of LCO 3.9.1 when DHR forced flow is stopped.
Note 2 allows one DHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no draining operation to further reduce RCS water level and that the capability exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the inoperable loop during a time when these tests are safe and possible.
An OPERABLE DHR loop consists of a DHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to DHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
Both DHR pumps may be aligned to the Refueling Water Storage Tank to support filling or draining the refueling cavity or for performance of required testing.
APPLICABILITY Two DHR loops are required to be OPERABLE, and one in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the DHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). DHR loop requirements in MODE 6, with the water level 23 ft above the top of the reactor vessel flange, are located in LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level."
ACTIONS A.1 and A.2 With fewer than the required loops OPERABLE, action shall be immediately initiated and continued until the DHR loop is restored to BWOG STS B 3.9.5-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 DHR and Coolant Circulation - Low Water Level B 3.9.5 BASES ACTIONS (continued)
- b. One door in each air lock must be closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.
With DHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most DHR problems and is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one DHR loop is in operation. The flow rate is determined by the flow rate necessary to provide efficient decay heat removal capability and to prevent thermal and boron stratification in the core.
In addition, during operation of the DHR loop with the water level in the vicinity of the reactor vessel nozzles, the DHR loop flow rate determination must also consider the DHR pump suction requirement.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the DHR System in the control room.
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
REFERENCES 1. FSAR, Section [ ] .
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWOG STS B 3.9.5-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 [ Verify the following valves are in the listed position 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ]
with power to the valve operator removed.
Number Position Function
[ ] [ ] [ ]
[ ] [ ] [ ]
[ ] [ ] [ ]
SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.3 [ Verify ECCS piping is full of water. 31 days ]
SR 3.5.2.43 Verify each ECCS pump's developed head at the In accordance test flow point is greater than or equal to the with the Inservice required developed head. Testing Program SR 3.5.2.54 Verify each ECCS automatic valve in the flow path [18] months that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
SR 3.5.2.65 Verify each ECCS pump starts automatically on an [18] months actual or simulated actuation signal.
SR 3.5.2.76 [ Verify, for each ECCS throttle valve listed below, [18] months ]
each position stop is in the correct position.
Valve Number
[ ]
[ ]
[ ]
WOG STS 3.5.2-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.87 Verify, by visual inspection, each ECCS train [18] months containment sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.
WOG STS 3.5.2-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE.
NOTE--------------------------------------------
An RHR train may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned to the ECCS mode of operation.
APPLICABILITY: MODE 4.
ACTIONS
NOTE-----------------------------------------------------------
LCO 3.0.4.b is not applicable to ECCS high head subsystem.
CONDITION REQUIRED ACTION COMPLETION TIME A. [ Required ECCS A.1 Initiate action to restore Immediately ]
residual heat removal required ECCS RHR (RHR) subsystem subsystem to OPERABLE inoperable. status.
B. Required ECCS [high B.1 Restore required ECCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> head subsystem] [high head subsystem] to inoperable. OPERABLE status.
C. Required Action and C.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> associated Completion Time [of Condition B] not met.
WOG STS 3.5.3-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all equipment In accordance required to be OPERABLE: with applicable SRs
[SR 3.5.2.1] [SR 3.5.2.76]
[SR 3.5.2.3] SR 3.5.2.87 SR 3.5.2.43 WOG STS 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued) stopped for a short period of time. The Note permits the stopping of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period is adequate to perform the test, and operating experience has shown that boron stratification is not a problem during this short period with no forced flow.
Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by initial startup test procedures:
- a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentrations less than required to meet SDM of LCO 3.1.1, therefore maintaining the margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation and
- b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.
Note 2 requires that the secondary side water temperature of each SG be
[50]°F above each of the RCS cold leg temperatures before the start of an RCP with any RCS cold leg temperature [275°F] [Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR]. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.
An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG, which has the minimum water level specified in SR 3.4.6.2.
Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
WOG STS B 3.4.6-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
WOG STS B 3.4.6-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO (continued)
Note 3 requires that the secondary side water temperature of each SG be
[50]°F above each of the RCS cold leg temperatures before the start of a reactor coolant pump (RCP) with an RCS cold leg temperature
[275°F] [Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR]. This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.
Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.
RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least [two] SGs is required to be [17]%.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2;"
LCO 3.4.5, "RCS Loops - MODE 3;"
LCO 3.4.6, "RCS Loops - MODE 4;"
LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled;"
LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -
High Water Level" (MODE 6), and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -
Low Water Level" (MODE 6).
WOG STS B 3.4.7-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required RHR pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. If secondary side water level is [17]% in at least two SGs, this Surveillance is not needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."
- 2. FSAR, Section [ ], ["Gas Management Program."]
WOG STS B 3.4.7-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO (continued)
Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.
An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger.
RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2,"
LCO 3.4.5, "RCS Loops - MODE 3,"
LCO 3.4.6, "RCS Loops - MODE 4,"
LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"
LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -
High Water Level" (MODE 6), and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -
Low Water Level" (MODE 6).
ACTIONS A.1 If one required RHR loop is inoperable, redundancy for RHR is lost.
Action must be initiated to restore a second loop to OPERABLE status.
The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
B.1 and B.2 If no required loop is OPERABLE or the required loop is not in operation, except during conditions permitted by Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status and operation. The required margin to WOG STS B 3.4.8-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must continue until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the required loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.
SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
WOG STS B 3.4.8-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES LCO (continued)
In MODES 1, 2, and 3, an ECCS train consists of a centrifugal charging subsystem, an SI subsystem, and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and automatically transferring suction to the containment sump.
During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 5).
The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.
As indicated in Note 1, the SI flow paths may be isolated for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room.
As indicated in Note 2, operation in MODE 3 with ECCS trains made incapable of injecting in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," is necessary for plants with an LTOP arming temperature at or near the MODE 3 boundary temperature of 350°F.
LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the LTOP arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the LTOP Applicability, and provide time to restore the inoperable pumps to OPERABLE status on exiting the LTOP Applicability.
APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA.
WOG STS B 3.5.2-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES APPLICABILITY (continued)
This LCO is only applicable in MODE 3 and above. Below MODE 3, the SI signal setpoint is manually bypassed by operator control, and system functional requirements are relaxed as described in LCO 3.5.3, "ECCS -
Shutdown."
In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops -
MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level."
ACTIONS A.1 With one or more trains inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on an NRC reliability evaluation (Ref. 56) and is a reasonable time for repair of many ECCS components.
An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available.
The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.
An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 56) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
WOG STS B 3.5.2-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES ACTIONS (continued)
Reference 67 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.
B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.1 Condition A is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available.
With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the facility is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power or by key locking the control in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned. These valves are of the type, described in Reference 76, that can disable the function of both ECCS trains and invalidate the accident analyses. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in view of other administrative controls that will ensure a mispositioned valve is unlikely.
WOG STS B 3.5.2-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.2.3 With the exception of the operating centrifugal charging pump, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases.
Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an SI signal or during shutdown cooling. The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the procedural controls governing system operation.
SR 3.5.2.43 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.
SRs are specified in the Inservice Testing Program of the ASME Code.
The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
WOG STS B 3.5.2-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.5 4 and SR 3.5.2.65 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.
SR 3.5.2.76 Realignment of valves in the flow path on an SI signal is necessary for proper ECCS performance. These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance is not required for plants with flow limiting orifices. The 18 month Frequency is based on the same reasons as those stated in SR 3.5.2.5 4 and SR 3.5.2.65.
SR 3.5.2.87 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to have access to the location, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience.
WOG STS B 3.5.2-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.
- 2. 10 CFR 50.46.
- 3. FSAR, Section [ ].
- 4. FSAR, Chapter [15], "Accident Analysis."
- 5. FSAR, Section [ ], ["Gas Management Program."]
- 56. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
WOG STS B 3.5.2-11 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications.
In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head).
The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.
APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also applies to this SAFETY Bases section.
ANALYSES Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA.
In MODE 4, an ECCS train consists of a centrifugal charging subsystem and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report.
During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to WOG STS B 3.5.3-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES LCO During a DBA, a minimum of one containment cooling train and one containment spray train are required to maintain the containment peak pressure and temperature below the design limits (Ref. 7). Additionally, one containment spray train is also required to remove iodine from the containment atmosphere and maintain concentrations below those assumed in the safety analysis. To ensure that these requirements are met, two containment spray trains and two containment cooling trains must be OPERABLE. Therefore, in the event of an accident, at least one train in each system operates, assuming the worst case single active failure occurs.
Each Containment Spray System typically includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an ESF actuation signal and automatically transferring suction to the containment sump.
Management of gas voids is important to Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 8).
Each Containment Cooling System typically includes demisters, cooling coils, dampers, fans, instruments, and controls to ensure an OPERABLE flow path.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and containment cooling trains.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray System and the Containment Cooling System are not required to be OPERABLE in MODES 5 and 6.
ACTIONS A.1 With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE spray and cooling trains are adequate to perform the iodine removal and containment cooling functions. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time takes into account the redundant heat removal capability afforded by the Containment Spray System, reasonable time for repairs, and low probability of a DBA occurring during this period.
WOG STS B 3.6.6A-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6A.3 Verifying that each [required] containment cooling train ESW cooling flow rate to each cooling unit is [700] gpm provides assurance that the design flow rate assumed in the safety analyses will be achieved (Ref. 3).
The Frequency was developed considering the known reliability of the Cooling Water System, the two train redundancy available, and the low probability of a significant degradation of flow occurring between surveillances.
SR 3.6.6A.4 Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 98). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program.
SR 3.6.6A.5 and SR 3.6.6A.6 These SRs require verification that each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation of a containment High-3 pressure signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power.
Operating experience has shown that these components usually pass the Surveillances when performed at the [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by SR 3.5.2.45. A single surveillance may be used to satisfy both requirements.
WOG STS B 3.6.6A-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6A.7 This SR requires verification that each [required] containment cooling train actuates upon receipt of an actual or simulated safety injection signal. The [18] month Frequency is based on engineering judgment and has been shown to be acceptable through operating experience. See SR 3.6.6A.5 and SR 3.6.6A.6, above, for further discussion of the basis for the [18] month Frequency.
SR 3.6.6A.8 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the passive design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the nozzles.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 3. FSAR, Section [ ].
- 4. FSAR, Section [ ].
- 5. FSAR, Section [ ].
- 6. FSAR, Section [ ].
- 7. FSAR, Section [ ].
- 8. FSAR, Section [ ], ["Gas Management Program."]
- 98. ASME Code for Operation and Maintenance of Nuclear Power Plants.
WOG STS B 3.6.6A-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES APPLICABLE SAFETY ANALYSES (continued)
The effect of an inadvertent containment spray actuation has been analyzed. An inadvertent spray actuation results in a [-2.0] psig containment pressure and is associated with the sudden cooling effect in the interior of the leak tight containment. Additional discussion is provided in the Bases for LCO 3.6.4A.
The modeled Containment Spray System actuation from the containment analysis is based upon a response time associated with exceeding the containment High-3 pressure setpoint to achieving full flow though the containment spray nozzles. The Containment Spray System total response time of [60] seconds includes diesel generator (DG) startup (for loss of offsite power), block loading of equipment, containment spray pump startup, and spray line filling (Ref. 3).
Containment cooling train performance for post accident conditions is given in Reference 4. The result of the analysis is that each train can provide 100% of the required peak cooling capacity during the post accident condition. The train post accident cooling capacity under varying containment ambient conditions, required to perform the accident analyses, is also shown in Reference 5.
The modeled Containment Cooling System actuation from the containment analysis is based on a response time associated with exceeding the containment High-3 pressure setpoint to achieving full Containment Cooling System air and safety grade cooling water flow.
The Containment Cooling System total response time of [60] seconds includes signal delay, DG startup (for loss of offsite power), and Service Water pump startup times (Ref. 6).
The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, a minimum of one containment cooling train and one containment spray train are required to maintain the containment peak pressure and temperature below the design limits (Ref. 7). To ensure that these requirements are met, two containment spray trains and two containment cooling units must be OPERABLE. Therefore, in the event of an accident, at least one train in each system operates, assuming the worst case single active failure occurs.
Each Containment Spray System typically includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an ESF actuation signal and automatically transferring suction to the containment sump.
WOG STS B 3.6.6B-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B Management of gas voids is important to Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 8).
BASES LCO (continued)
Each Containment Cooling System typically includes demisters, cooling coils, dampers, instruments, and controls to ensure an OPERABLE flow path.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and containment cooling trains.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray System and the Containment Cooling System are not required to be OPERABLE in MODES 5 and 6.
ACTIONS A.1 If one containment spray train is inoperable, it must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing at least 100% of the heat removal needs (for the condition of one containment spray train inoperable) after an accident. The 7 day Completion Time was chosen in light of the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System and the low probability of DBA occurring during this period.
B.1 If one of the required containment cooling trains is inoperable, it must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing at least 100% of the heat removal needs (for the Condition of one containment cooling train inoperable) after an accident. The 7 day Completion Time was chosen based on the same reasons as given in Required Action A.1.
C.1 With two of the required containment spray trains inoperable, one must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded condition are capable of providing at least 100% of the heat removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was WOG STS B 3.6.6B-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES SURVEILLANCE SR 3.6.6B.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System flow path provides assurance that the proper flow path exists for Containment Spray System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct positions prior to being secured. This SR does not require testing or valve manipulation. Rather, it involves verification that those valves outside containment (only check valves are inside containment) and capable of potentially being mispositioned are in the correct position.
SR 3.6.6B.2 Operating each [required] containment cooling train fan unit for 15 minutes ensures that all trains are OPERABLE and all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.
The 31 day Frequency was developed based on the known reliability of the fan units and controls, the two train redundancy available, and the low probability of significant degradation of the containment cooling train occurring between surveillances.
SR 3.6.6B.3 Verifying that each [required] containment cooling train ESW cooling flow rate to each cooling unit is [700] gpm provides assurance that the design flow rate assumed in the analyses will be achieved (Ref. 3). The Frequency was developed considering the known reliability of the Cooling Water System, the two train redundancy available, and the low probability of a significant degradation of flow occurring between surveillances.
SR 3.6.6B.4 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 98). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice WOG STS B 3.6.6B-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES SURVEILLANCE REQUIREMENTS (continued) inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
SR 3.6.6B.5 and SR 3.6.6B.6 These SRs require verification that each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated containment High-3 pressure signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillances when performed at the [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by SR 3.5.2.45. A single surveillance may be used to satisfy both requirements.
SR 3.6.6B.7 This SR ensures that each [required] containment cooling train actuates upon receipt of an actual or simulated safety injection signal. The
[18] month Frequency is based on engineering judgment and has been proven acceptable through operating experience. See SR 3.6.6B.5 and SR 3.6.6B.6, above, for further discussion of the basis for the [18] month Frequency.
WOG STS B 3.6.6B-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6B.8 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. Because of the passive design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the spray nozzles.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 3. FSAR, Section [15].
- 4. FSAR, Section [6.2].
- 5. FSAR, Section [ ].
- 6. FSAR, Section [ ].
- 7. FSAR, Section [ ].
- 8. FSAR, Section [ ], ["Gas Management Program."]
- 89. ASME Code for Operation and Maintenance of Nuclear Power Plants.
WOG STS B 3.6.6B-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray System (Ice Condenser)
B 3.6.6C BASES APPLICABLE SAFETY ANALYSES (continued)
The modeled Containment Spray System actuation from the containment analysis is based on a response time associated with exceeding the containment High-3 pressure signal setpoint to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time of [45] seconds is composed of signal delay, diesel generator startup, and system startup time.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 4).
Inadvertent actuation of the Containment Spray System is evaluated in the analysis, and the resultant reduction in containment pressure is calculated. The maximum calculated reduction in containment pressure resulted in a containment external pressure load of [1.2] psid, which is below the containment design external pressure load.
The Containment Spray System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, one train of Containment Spray System is required to provide the heat removal capability assumed in the safety analyses.
Additionally, a minimum of one train of the Containment Spray System, with spray pH adjusted by the Spray Additive System, is required to scavenge iodine fission products from the containment atmosphere and ensure their retention in the containment sump water. To ensure that these requirements are met, two containment spray trains must be OPERABLE with power from two safety related, independent power supplies. Therefore, in the event of an accident, at least one train in each system operates.
Each Containment Spray System typically includes a spray pump, headers, valves, heat enhancers, nozzles, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an ESF actuation signal and automatically transferring suction to the containment sump.
WOG STS B 3.6.6C-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray System (Ice Condenser)
B 3.6.6C Management of gas voids is important to Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 5).
BASES APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the Containment Spray System.
In MODES 5 and 6, the probability and consequences of these events are reduced because of the pressure and temperature limitations of these MODES. Thus, the Containment Spray System is not required to be OPERABLE in MODE 5 or 6.
ACTIONS A.1 With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.
B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.
SURVEILLANCE SR 3.6.6C.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR does not require any testing or valve manipulation. Rather, it involves verification that WOG STS B 3.6.6C-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray System (Ice Condenser)
B 3.6.6C BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 65). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillances when performed at the
[18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by SR 3.6.6.3. A single surveillance may be used to satisfy both requirements.
SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. Because of the passive design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the spray nozzles.
WOG STS B 3.6.6C-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray System (Ice Condenser)
B 3.6.6C BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 2. FSAR, Section [6.2].
- 3. 10 CFR 50.49.
- 5. FSAR, Section [ ], ["Gas Management Program."]
- 65. ASME Code for Operation and Maintenance of Nuclear Power Plants.
WOG STS B 3.6.6C-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 QS System (Subatmospheric)
B 3.6.6D BASES APPLICABLE SAFETY ANALYSES (continued)
For certain aspects of accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 3).
Inadvertent actuation of the QS System is evaluated in the analysis, and the resultant reduction in containment pressure is calculated. The maximum calculated reduction in containment pressure resulted in a containment external pressure load of [unit specific pressure], which is below the containment design external pressure load.
The QS System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, one train of the QS System is required to provide the heat removal capability assumed in the safety analyses for containment. In addition, one QS System train, with spray pH adjusted by the Spray Additive System, is required to scavenge iodine fission products from the containment atmosphere and ensure their retention in the containment sump water. To ensure that these requirements are met, two QS System trains must be OPERABLE with power from two safety related, independent power supplies. Therefore, in the event of an accident, at least one train in each system will operate, assuming that the worst case single active failure occurs.
Each QS System includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST.
Management of gas voids is important to QS System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 4).
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the QS System.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the QS System is not required to be OPERABLE in MODE 5 or 6.
WOG STS B 3.6.6D-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 QS System (Subatmospheric)
B 3.6.6D BASES ACTIONS A.1 If one QS train is inoperable, it must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.
B.1 and B.2 If the Required Action and associated Completion Time are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.6D.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the QS System provides assurance that the proper flow path exists for QS System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they were verified to be in the correct position prior to being secured. This SR does not require any testing or valve manipulation. Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position.
SR 3.6.6D.2 Verifying that each QS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that QS pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 45). Since the QS System pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
WOG STS B 3.6.6D-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 QS System (Subatmospheric)
B 3.6.6D BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6D.3 and SR 3.6.6D.4 These SRs ensure that each QS automatic valve actuates to its correct position and each QS pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power.
Operating experience has shown that these components usually pass the Surveillances when performed at an [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.6.6D.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. Due to the passive nature of the design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the nozzles.
REFERENCES 1. FSAR, Section [6.2].
- 2. 10 CFR 50.49.
- 4. FSAR, Section [ ], ["Gas Management Program."]
- 45. ASME Code for Operation and Maintenance of Nuclear Power Plants.
WOG STS B 3.6.6D-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RS System (Subatmospheric)
B 3.6.6E BASES APPLICABLE SAFETY ANALYSES (continued) backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 3).
The RS System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, one train (two subsystems) of the RS System is required to provide the minimum heat removal capability assumed in the safety analysis. To ensure that this requirement is met, four RS subsystems
[and a casing cooling tank] must be OPERABLE. This will ensure that at least one train will operate assuming the worst case single failure occurs, which is in the ESF power supply.
Management of gas voids is important to RS System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 4).
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the RS System.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the RS System is not required to be OPERABLE in MODE 5 or 6.
ACTIONS A.1 With one of the RS subsystems inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing at least 100% of the heat removal needs (i.e., 150% when one RS subsystem is inoperable) after an accident. The 7 day Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the RS and QS systems and the low probability of a DBA occurring during this period.
B.1 With two of the required RS subsystems inoperable in the same train, at least one of the inoperable RS subsystems must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded condition are capable of providing 100% of the heat removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into WOG STS B 3.6.6E-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RS System (Subatmospheric)
B 3.6.6E BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6E.4 Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the RS System and casing cooling tank provides assurance that the proper flow path exists for operation of the RS System. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified as being in the correct position prior to being secured. This SR does not require any testing or valve manipulation. Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position.
SR 3.6.6E.5 Verifying that each RS [and casing cooling] pump's developed head at the flow test point is greater than or equal to the required developed head ensures that these pumps' performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 54). Since the QS System pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
SR 3.6.6E.6 These SRs ensure that each automatic valve actuates and that the RS System and casing cooling pumps start upon receipt of an actual or simulated High-High containment pressure signal. Start delay times are also verified for the RS System pumps. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency. Therefore, the Frequency was considered to be acceptable from a reliability standpoint.
WOG STS B 3.6.6E-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RS System (Subatmospheric)
B 3.6.6E BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6E.7 This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment will meet its design bases objective. An air or smoke test is performed through each spray header. Due to the passive design of the spray header and its normally dry state, a test at
[the first refueling and at] 10 year intervals is considered adequate for detecting obstruction of the nozzles.
REFERENCES 1. FSAR, Section [6.2].
- 2. 10 CFR 50.49.
- 4. FSAR, Section [ ], ["Gas Management Program."]
- 54. ASME Code for Operation and Maintenance of Nuclear Power Plants.
WOG STS B 3.6.6E-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Spray Additive System (Atmospheric, Subatmospheric, Ice Condenser, and Dual)
B 3.6.7 BASES LCO (continued) spray flow until the Containment Spray System suction path is switched from the RWST to the containment sump, and to raise the average spray solution pH to a level conducive to iodine removal, namely, to between [7.2 and 11.0]. This pH range maximizes the effectiveness of the iodine removal mechanism without introducing conditions that may induce caustic stress corrosion cracking of mechanical system components. In addition, it is essential that valves in the Spray Additive System flow paths are properly positioned and that automatic valves are capable of activating to their correct positions.
Management of gas voids is important to Spray Additive System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment requiring the operation of the Spray Additive System. The Spray Additive System assists in reducing the iodine fission product inventory prior to release to the environment.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Thus, the Spray Additive System is not required to be OPERABLE in MODE 5 or 6.
ACTIONS A.1 If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The pH adjustment of the Containment Spray System flow for corrosion protection and iodine removal enhancement is reduced in this condition. The Containment Spray System would still be available and would remove some iodine from the containment atmosphere in the event of a DBA. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period.
B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in WOG STS B 3.6.7-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Spray Additive System (Atmospheric, Subatmospheric, Ice Condenser, and Dual)
B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the
[18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the Containment Spray System, the flow rate in the Spray Additive System is verified once every 5 years. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation. Due to the passive nature of the spray additive flow controls, the 5 year Frequency is sufficient to identify component degradation that may affect flow rate.
REFERENCES 1. FSAR, Chapter [15].
- 2. FSAR, Section [ ], ["Gas Management Program."]
WOG STS B 3.6.7-6 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR and Coolant Circulation - High Water Level B 3.9.5 BASES LCO (continued)
- b. Mixing of borated coolant to minimize the possibility of criticality, and
- c. Indication of reactor coolant temperature.
An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
The LCO is modified by a Note that allows the required operating RHR loop to be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1. Boron concentration reduction with coolant at boron concentrations less than required to assure the RCS boron concentration is maintained is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing.
During this 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.
APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, with the water level 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LCO 3.9.7, "Refueling Cavity Water Level." Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level < 23 ft are located in LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level."
WOG STS B 3.9.5-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR and Coolant Circulation - High Water Level B 3.9.5 BASES ACTIONS (continued)
- b. One door in each air lock must be closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.
With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the RHR System.
REFERENCES 1. FSAR, Section [5.5.7].
- 2. FSAR, Section [ ], ["Gas Management Program."]
WOG STS B 3.9.5-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES LCO (continued)
This LCO is modified by two Notes. Note 1 permits the RHR pumps to be removed from operation for 15 minutes when switching from one train to another. The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short [and the core outlet temperature is maintained > 10 degrees F below saturation temperature].
The Note prohibits boron dilution or draining operations when RHR forced flow is stopped.
Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no draining operation to further reduce RCS water level and that the capability exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the inoperable loop during a time when these tests are safe and possible.
An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
Both RHR pumps may be aligned to the Refueling Water Storage Tank to support filling or draining the refueling cavity or for performance of required testing.
APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level 23 ft are located in LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level."
WOG STS B 3.9.6-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES ACTIONS (continued)
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.
With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room.
Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
REFERENCES 1. FSAR, Section [5.5.7].
- 2. FSAR, Section [ ], ["Gas Management Program."]
WOG STS B 3.9.6-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Less than 100% of the D.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE train available.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 [ Verify the following valves are in the listed position 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ]
with power to the valve operator removed [and key locked in position].
Valve Number Position Function
[ ] [ ] [ ]
[ ] [ ] [ ]
[ ] [ ] [ ]
SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.3 [ Verify ECCS piping is full of water. 31 days ]
SR 3.5.2.43 Verify each ECCS pump's developed head at the In accordance test flow point is greater than or equal to the with the Inservice required developed head. Testing Program SR 3.5.2.54 [ Verify each charging pump develops a flow of In accordance
[36] gpm at a discharge pressure of [2200] psig. with the Inservice Testing Program ]
CEOG STS 3.5.2-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.65 Verify each ECCS automatic valve that is not [18] months locked, sealed, or otherwise secured in position, in the flow path actuates to the correct position on an actual or simulated actuation signal.
SR 3.5.2.76 Verify each ECCS pump starts automatically on an [18] months actual or simulated actuation signal.
SR 3.5.2.87 Verify each LPSI pump stops on an actual or [18] months simulated actuation signal.
SR 3.5.2.98 [ Verify, for each ECCS throttle valve listed below, [18] months ]
each position stop is in the correct position.
Valve Number
[ ]
[ ]
SR 3.5.2.109 Verify, by visual inspection, each ECCS train [18] months containment sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.
CEOG STS 3.5.2-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS - Shutdown LCO 3.5.3 One high pressure safety injection (HPSI) train shall be OPERABLE.
APPLICABILITY: MODE 3 with pressurizer pressure < [1700] psia, MODE 4.
ACTIONS
NOTE-----------------------------------------------------------
LCO 3.0.4.b is not applicable to ECCS High Pressure Safety Injection subsystem when entering MODE 4.
CONDITION REQUIRED ACTION COMPLETION TIME A. Required HPSI train A.1 Restore required HPSI train 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable. to OPERABLE status.
B. Required Action and B.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable: In accordance with applicable
[SR 3.5.2.1] SR 3.5.2.65 SRs SR 3.5.2.2 SR 3.5.2.76
[SR 3.5.2.3] [SR 3.5.2.98]
SR 3.5.2.43 SR 3.5.2.109 CEOG STS 3.5.3-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued) pumps are stopped. As decay heat diminishes, the effects on RCS temperature and pressure diminish. Without cooling by forced flow, higher heat loads will cause the reactor coolant temperature and pressure to increase at a rate proportional to the decay heat load. Because pressure can increase, the applicable system pressure limits (pressure and temperature (P/T) limits or low temperature overpressure protection (LTOP) limits) must be observed and forced SDC flow or heat removal via the SGs must be re-established prior to reaching the pressure limit. The circumstances for stopping both RCPs or SDC pumps are to be limited to situations where:
- a. Pressure and temperature increases can be maintained well within the allowable pressure (P/T limits and LTOP) and 10°F subcooling limits or
- b. An alternate heat removal path through the SGs is in operation.
Note 2 requires that either of the following two conditions be satisfied before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR:
- a. Pressurizer water level is < [60]% or
- b. Secondary side water temperature in each SG is < [100]°F above each of the RCS cold leg temperatures.
Satisfying either of the above conditions will preclude a large pressure surge in the RCS when the RCP is started.
An OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE and has the minimum water level specified in SR 3.4.6.2.
Similarly, for the SDC System, an OPERABLE SDC train is composed of the OPERABLE SDC pump(s) capable of providing forced flow to the SDC heat exchanger(s). RCPs and SDC pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
CEOG STS B 3.4.6-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS loop or SDC train can be placed in operation, if needed to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
CEOG STS B 3.4.6-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO (continued)
Note 3 requires that either of the following two conditions be satisfied before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR:
- a. Pressurizer water level must be < [60]% or
- b. Secondary side water temperature in each SG must be < [100]°F above each of the RCS cold leg temperatures.
Satisfying either of the above conditions will preclude a low temperature overpressure event due to a thermal transient when the RCP is started.
Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting SDC trains to not be in operation when at least one RCP is in operation. This Note provides for the transition to MODE 4 where an RCP is permitted to be in operation and replaces the RCS circulation function provided by the SDC trains.
An OPERABLE SDC train is composed of an OPERABLE SDC pump and an OPERABLE SDC heat exchanger.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
SDC pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE.
APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation to remove decay heat from the core and to provide proper boron mixing.
One SDC train provides sufficient circulation for these purposes.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2,"
LCO 3.4.5, "RCS Loops - MODE 3,"
LCO 3.4.6, "RCS Loops - MODE 4,"
LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled,"
LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level" (MODE 6), and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level" (MODE 6).
CEOG STS B 3.4.7-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.7.2 Verifying the SGs are OPERABLE by ensuring their secondary side water levels are [25%] ensures that redundant heat removal paths are available if the second SDC train is inoperable. The Surveillance is required to be performed when the LCO requirement is being met by use of the SGs. If both SDC trains are OPERABLE, this SR is not needed.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions.
SR 3.4.7.3 Verification that each required SDC train is OPERABLE ensures that redundant paths for decay heat removal are available. The requirement also ensures that the additional train can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability.
The Surveillance is required to be performed when the LCO requirement is being met by one of two SDC trains, e.g., both SGs have < [25]% water level. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."
- 2. FSAR, Section [ ], ["Gas Management Program."]
CEOG STS B 3.4.7-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO (continued)
An OPERABLE SDC train is composed of an OPERABLE SDC pump capable of providing forced flow to an OPERABLE SDC heat exchanger, along with the appropriate flow and temperature instrumentation for control, protection, and indication. SDC pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the SDC System.
Operation in other MODES is covered by:
LCO 3.4.4, "RCS Loops - MODES 1 and 2,"
LCO 3.4.5, "RCS Loops - MODE 3,"
LCO 3.4.6, "RCS Loops - MODE 4,"
LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"
LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level" (MODE 6), and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level" (MODE 6).
ACTIONS A.1 If one required SDC train is inoperable, redundancy for heat removal is lost. Action must be initiated immediately to restore a second train to OPERABLE status. The Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
B.1 and B.2 If no required SDC train is OPERABLE or the required train is not in operation, except as provided in Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended.
Action to restore one SDC train to OPERABLE status and operation must be initiated immediately. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable CEOG STS B 3.4.8-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the required SDC train is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing decay heat removal. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency has been shown by operating practice to be sufficient to regularly assess degradation and verify operation is within safety analyses assumptions.
SR 3.4.8.2 Verification that each required train is OPERABLE ensures that redundant paths for heat removal are available and that an additional train can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and indicated power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a required pump is not in operation.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
CEOG STS B 3.4.8-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES APPLICABLE SAFETY ANALYSES (continued)
The LCO ensures that an ECCS train will deliver sufficient water to match decay heat boiloff rates soon enough to minimize core uncovery for a large LOCA. It also ensures that the HPSI pump will deliver sufficient water during a small break LOCA and provide sufficient boron to maintain the core subcritical following an SLB. For smaller LOCAs, the charging pumps deliver sufficient fluid to maintain RCS inventory until the RCS can be depressurized below the HPSI pumps' shutoff head. During this period of a small break LOCA, the SGs continue to serve as the heat sink providing core cooling.
ECCS - Operating satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO In MODES 1, 2, and 3, with pressurizer pressure 1700 psia, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming there is a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.
In MODES 1 and 2, and in MODE 3 with pressurizer pressure 1700 psia, an ECCS train consists of an HPSI subsystem, an LPSI subsystem, and a charging pump.
Each train includes the piping, instruments, and controls to ensure the availability of an OPERABLE flow path capable of taking suction from the RWT on an SIAS and automatically transferring suction to the containment sump upon a recirculation actuation signal (RAS).
During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the RWT to the RCS, via the HPSI and LPSI pumps and their respective supply headers, to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply part of its flow to the RCS hot legs via the shutdown cooling (SDC) suction nozzles. The charging pump flow path takes suction from the RWT and supplies the RCS via the normal charging lines.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 4).
The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.
CEOG STS B 3.5.2-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure 1700 psia, the ECCS OPERABILITY requirements for the limiting Design Basis Accident (DBA) large break LOCA are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the pump performance curve and has less dependence on power. The charging pump performance requirements are based on a small break LOCA. The requirements of MODES 2 and 3, with RCS pressure 1700 psia, are bounded by the MODE 1 analysis.
The ECCS functional requirements of MODE 3, with RCS pressure
< 1700 psia, and MODE 4 are described in LCO 3.5.3, "ECCS -
Shutdown."
In MODES 5 and 6, unit conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops -
MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level."
ACTIONS A.1 With one LPSI subsystem inoperable, action must be taken to restore OPERABLE status within 7 days. In this condition, the remaining OPERABLE ECCS train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure to the remaining LPSI subsystem could result in loss of ECCS function. The 7 day Completion Time is reasonable to perform corrective maintenance on the inoperable LPSI subsystem. The 7 day Completion Time is based on the findings of the deterministic and probabilistic analysis in Reference 67. Reference 6 7 concluded that extending the Completion Time to 7 days for an inoperable LPSI train provides plant operational flexibility while simultaneously reducing overall plant risk.
This is because the risks incurred by having the LPSI train unavailable for a longer time at power will be substantially offset by the benefits associated with avoiding unnecessary plant transitions and by reducing risk during plant shutdown operations.
CEOG STS B 3.5.2-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES ACTIONS (continued)
B.1 If one or more trains are inoperable except for reasons other than Condition A (one LPSI subsystem inoperable) and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available, the inoperable components must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on an NRC study (Ref. 45) using a reliability evaluation and is a reasonable amount of time to effect many repairs.
An ECCS train is inoperable if it is not capable of delivering the design flow to the RCS. The individual components are inoperable if they are not capable of performing their design function, or if supporting systems are not available.
The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. This allows increased flexibility in plant operations when components in opposite trains are inoperable.
An event accompanied by a loss of offsite power and the failure of an emergency DG can disable one ECCS train until power is restored. A reliability analysis (Ref. 46) has shown that the impact with one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Reference 65 describes situations in which one component, such as a shutdown cooling total flow control valve, can disable both ECCS trains.
With one or more components inoperable, such that 100% of the equivalent flow to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be immediately entered.
CEOG STS B 3.5.2-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES ACTIONS (continued)
C.1 and C.2 If the inoperable train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to < 1700 psia within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems.
D.1 Condition B is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available.
With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the facility is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removing power or by key locking the control in the correct position ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 65, which can disable the function of both ECCS trains and invalidate the accident analysis. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in view of other administrative controls ensuring that a mispositioned valve is an unlikely possibility.
SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, CEOG STS B 3.5.2-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position.
The 31 day Frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.2.3 With the exception of systems in operation, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases.
Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an SIAS or during SDC. The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the adequacy of the procedural controls governing system operation.
SR 3.5.2.43 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis.
SRs are specified in the Inservice Testing Program of the ASME Code.
The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
CEOG STS B 3.5.2-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.54 Discharge head at design flow is a normal test of charging pump performance required by the ASME Code. A quarterly Frequency for such tests is a Code requirement. Such inservice inspections detect component degradation and incipient failures.
SR 3.5.2.65, SR 3.5.2.76, and SR 3.5.2.87 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SIAS and on an RAS, that each ECCS pump starts on receipt of an actual or simulated SIAS, and that the LPSI pumps stop on receipt of an actual or simulated RAS. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, and equipment performance is monitored as part of the Inservice Testing Program.
SR 3.5.2.98 Realignment of valves in the flow path on an SIAS is necessary for proper ECCS performance. The safety injection valves have stops to position them properly so that flow is restricted to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This SR is not required for units with flow limiting orifices. The 18 month Frequency is based on the same factors as those stated above for SR 3.5.2.65, SR 3.5.2.76, and SR 3.5.2.87.
CEOG STS B 3.5.2-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.109 Periodic inspection of the containment sump ensures that it is unrestricted and stays in proper operating condition. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during an outage, on the need to have access to the location, and on the potential for unplanned transients if the Surveillance were performed with the reactor at power. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.
- 2. 10 CFR 50.46.
- 3. FSAR, Chapter [6].
- 4. FSAR, Section [ ], ["Gas Management Program."]
- 54. NRC Memorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
- 65. IE Information Notice No. 87-01, January 6, 1987.
CEOG STS B 3.5.2-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
B 3.5.3 ECCS - Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2, "ECCS - Operating," is applicable to these Bases, with the following modifications.
In MODE 3 with pressurizer pressure < 1700 psia and in MODE 4, an ECCS train is defined as one high pressure safety injection (HPSI) subsystem. The HPSI flow path consists of piping, valves, and pumps that enable water from the refueling water tank (RWT) to be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.
APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 is applicable to SAFETY these Bases.
ANALYSES Due to the stable conditions associated with operation in MODE 4, and the reduced probability of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. Included in these reductions is that certain automatic safety injection (SI) actuation signals are not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
Only one train of ECCS is required for MODE 4. Protection against single failures is not relied on for this MODE of operation.
ECCS - Shutdown satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO In MODE 3 with pressurizer pressure < 1700 psia, an ECCS subsystem is composed of a single HPSI subsystem. Each HPSI subsystem includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWT and transferring suction to the containment sump.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report.
During an event requiring ECCS actuation, a flow path is required to supply water from the RWT to the RCS via the HPSI pumps and their respective supply headers to each of the four cold leg injection nozzles.
In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs.
CEOG STS B 3.5.3-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES APPLICABLE SAFETY ANALYSES (continued)
The modeled Containment Spray System actuation from the containment analysis is based upon a response time associated with exceeding the containment High-High pressure setpoint coincident with an SIAS to achieve full flow through the containment spray nozzles. The Containment Spray System total response time of [60] seconds includes diesel generator startup (for loss of offsite power), block loading of equipment, containment spray pump startup, and spray line filling (Ref. 2).
The performance of the containment cooling train for post accident conditions is given in Reference 3. The result of the analysis is that each train can provide 50% of the required peak cooling capacity during the post accident condition. The train post accident cooling capacity under varying containment ambient conditions, required to perform the accident analyses, is also shown in Reference 4.
The modeled Containment Cooling System actuation from the containment analysis is based upon the unit specific response time associated with exceeding the CCAS to achieve full Containment Cooling System air and safety grade cooling water flow.
The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, a minimum of two containment cooling trains or two containment spray trains, or one of each, is required to maintain the containment peak pressure and temperature below the design limits (Ref. 5). Additionally, one containment spray train is also required to remove iodine from the containment atmosphere and maintain concentrations below those assumed in the safety analysis. To ensure that these requirements are met, two containment spray trains and two containment cooling units must be OPERABLE. Therefore, in the event of an accident, the minimum requirements are met, assuming that the worst case single active failure occurs.
Each Containment Spray System includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWT upon an ESF actuation signal and automatically transferring suction to the containment sump.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 6).
CEOG STS B 3.6.6A-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature, requiring the operation of the containment spray trains and containment cooling trains.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray and Containment Cooling systems are not required to be OPERABLE in MODES 5 and 6.
ACTIONS A.1
REVIEWERS NOTE-----------------------------------
Utilization of the 7 day Completion Time for Required Action A.1 is dependent on the licensee adopting CE NPSD-1045-A (Ref. 76) and meeting the requirements of the Topical Report and the associated Safety Evaluation. Otherwise, a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time applies.
With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within [7] days. In this Condition, the remaining OPERABLE spray and cooling trains are adequate to perform the iodine removal and containment cooling functions. The [7] day Completion Time takes into account the redundant heat removal capability afforded by the Containment Spray System, reasonable time for repairs, and the findings of Ref. 76.
B.1 and B.2 If the inoperable containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time for the restoration of the containment spray train and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.
CEOG STS B 3.6.6A-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6A.6 and SR 3.6.6A.7 These SRs verify that each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillances when performed at the [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by SR 3.5.2.45. A single surveillance may be used to satisfy both requirements.
SR 3.6.6A.8 This SR verifies that each containment cooling train actuates upon receipt of an actual or simulated actuation signal. The [18] month Frequency is based on engineering judgment and has been shown to be acceptable through operating experience. See SR 3.6.6A.6 and SR 3.6.6A.7, above, for further discussion of the basis for the [18] month Frequency.
SR 3.6.6A.9 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the passive design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the spray nozzles.
CEOG STS B 3.6.6A-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6A BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 2. FSAR, Section [ ].
- 3. FSAR, Section [ ].
- 4. FSAR, Section [ ].
- 5. FSAR, Section [ ].
- 6. FSAR, Section [ ], ["Gas Management Program."]
- 76. CE NPSD-1045-A, CEOG Joint Application Report for Modification to the Containment Spray System Technical Specifications,"
March 2000.
- 87. ASME Code for Operation and Maintenance of Nuclear Power Plants.
CEOG STS B 3.6.6A-11 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES APPLICABLE SAFETY ANALYSES (continued)
The modeled Containment Cooling System actuation from the containment analysis is based on the unit specific response time associated with exceeding the CCAS to achieve full Containment Cooling System air and safety grade cooling water flow.
The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO During a DBA, a minimum of two containment cooling trains or two containment spray trains, or one of each, is required to maintain the containment peak pressure and temperature below the design limits (Ref. 5). To ensure that these requirements are met, two containment spray trains and two containment cooling units must be OPERABLE.
Therefore, in the event of an accident, the minimum requirements are met, assuming the worst case single active failure occurs.
Each Containment Spray System typically includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWT upon an ESF actuation signal and automatically transferring suction to the containment sump.
Management of gas voids is important to Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 6).
Each Containment Cooling System typically includes demisters, cooling coils, dampers, fans, instruments, and controls to ensure an OPERABLE flow path.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and containment cooling trains.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray and Containment Cooling systems are not required to be OPERABLE in MODES 5 and 6.
ACTIONS A.1 With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing greater CEOG STS B 3.6.6B-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6B.2 Operating each containment cooling train fan unit for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring between surveillances.
SR 3.6.6B.3 Verifying a service water flow rate of [2000] gpm to each cooling unit provides assurance the design flow rate assumed in the safety analyses will be achieved (Ref. 2). Also considered in selecting this Frequency were the known reliability of the cooling water system, the two train redundancy, and the low probability of a significant degradation of flow occurring between surveillances.
[ SR 3.6.6B.4 Verifying the containment spray header is full of water to the [100] ft level minimizes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containment analysis. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances. ]
SR 3.6.6B.5 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 76). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump CEOG STS B 3.6.6B-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES SURVEILLANCE REQUIREMENTS (continued) design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
SR 3.6.6B.6 and SR 3.6.6B.7 These SRs verify each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillances when performed at the [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by SR 3.5.2.45. A single surveillance may be used to satisfy both requirements.
SR 3.6.6B.8 This SR verifies each containment cooling train actuates upon receipt of an actual or simulated actuation signal. The [18] month Frequency is based on engineering judgment and has been shown to be acceptable through operating experience. See SR 3.6.6B.6 and SR 3.6.6B.7, above, for further discussion of the basis for the [18] month Frequency.
SR 3.6.6B.9 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the passive design of the nozzle, a test at [the first refueling and at] 10 year intervals is considered adequate to detect obstruction of the spray nozzles.
CEOG STS B 3.6.6B-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Containment Spray and Cooling Systems (Atmospheric and Dual)
B 3.6.6B BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.
- 2. FSAR, Section [ ].
- 3. FSAR, Sections [ ].
- 4. FSAR, Section [ ].
- 5. FSAR, Section [ ].
- 6. FSAR, Section [ ], ["Gas Management Program."]
- 76. ASME Code for Operation and Maintenance of Nuclear Power Plants.
CEOG STS B 3.6.6B-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 Spray Additive System (Atmospheric and Dual)
B 3.6.7 BASES APPLICABLE SAFETY ANALYSES (continued)
The DBA response time assumed for the Spray Additive System is the same as for the Containment Spray System and is discussed in the Bases for Specification 3.6.6, "Containment Spray and Cooling Systems."
The DBA analyses assume that one train of the Containment Spray System/Spray Additive System is inoperable and that the entire spray additive tank volume is added to the remaining Containment Spray System flow path.
During a LOCA, the iodine inventory released to the containment is considered to be released instantaneously and uniformly distributed in the containment free volume. The containment volume is made up of sprayed and unsprayed regions. The sprayed region is enveloped by direct spray and mixed by the dome air circulators and emergency fan coolers. Mixing between the sprayed and unsprayed regions is facilitated by the emergency fan coolers and condensation of steam by the sprays.
The Spray Additive System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO The Spray Additive System is necessary to reduce the release of radioactive material to the environment in the event of a DBA. To be considered OPERABLE, the volume and concentration of the spray additive solution must be sufficient to maintain the pH of the spray solution between [9.0 and 10.0] in the injection mode and [8.0 and 9.0] in the recirculation mode. This pH range maximizes the effectiveness of the iodine removal mechanism, without introducing conditions that may induce caustic stress corrosion cracking of mechanical components.
Management of gas voids is important to Spray Additive System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
During a LOCA, one Spray Additive System train is capable of providing 100% of the required iodine removal capacity. To ensure at least one train is available in the event of the limiting single failure, both trains must be maintained in an OPERABLE status.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment requiring the operation of the Spray Additive System. The Spray Additive System assists in reducing the iodine fission product inventory prior to release to the environment.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Thus, the Spray Additive System is not required to be OPERABLE in MODES 5 and 6.
CEOG STS B 3.6.7-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 Spray Additive System (Atmospheric and Dual)
B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.7.5 This SR verifies that each automatic valve in the Spray Additive System flow path actuates to its correct position on a CSAS. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
[ SR 3.6.7.6 To ensure that the correct pH level is established in the borated water solution provided by the Containment Spray System, the flow rate in the Spray Additive System is verified once per 5 years. This SR provides assurance that the correct amount of N2H4 will be metered into the flow path upon Containment Spray System initiation. Due to the passive nature of the spray additive flow controls, the 5 year Frequency is sufficient to identify component degradation that may affect flow rate. ]
REFERENCES 1. FSAR, Section [ ].
- 2. FSAR, Section [ ], ["Gas Management Program."]
CEOG STS B 3.6.7-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 SDC and Coolant Circulation - High Water Level B 3.9.4 BASES LCO Only one SDC loop is required for decay heat removal in MODE 6, with water level 23 ft above the top of the reactor vessel flange. Only one SDC loop is required because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one SDC loop must be OPERABLE and in operation to provide:
- a. Removal of decay heat,
- b. Mixing of borated coolant to minimize the possibility of a criticality, and
- c. Indication of reactor coolant temperature.
An OPERABLE SDC loop includes an SDC pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
Both SDC pumps may be aligned to the Refueling Water Storage Tank to support filing or draining the refueling cavity or for performance of required testing.
The LCO is modified by a Note that allows the required operating SDC loop to be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in each 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, provided no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1. Boron concentration reduction with coolant at boron concentrations less than required to assure the RCS boron concentration is maintained is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles, and RCS to SDC isolation valve testing.
During this 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.
APPLICABILITY One SDC loop must be in operation in MODE 6, with the water level 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft level was selected because it corresponds to the 23 ft requirement established for fuel movement in LCO 3.9.6, "Refueling Water Level." Requirements for the SDC System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). SDC loop requirements in MODE 6, with the water level < 23 ft above the top of the CEOG STS B 3.9.4-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 SDC and Coolant Circulation - High Water Level B 3.9.4 BASES ACTIONS (continued)
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.
With SDC loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most SDC problems and is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the SDC loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the SDC System.
REFERENCES 1. FSAR, Section [ ].
- 2. FSAR, Section [ ], ["Gas Management Program."]
CEOG STS B 3.9.4-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 SDC and Coolant Circulation - Low Water Level B 3.9.5 BASES LCO (continued)
- c. Indication of reactor coolant temperature.
This LCO is modified by two Notes. Note 1 permits the SDC pumps to be removed from operation for 15 minutes when switching from one train to another. The circumstances for stopping both SDC pumps are to be limited to situations when the outage time is short [and the core outlet temperature is maintained > 10 degrees F below saturation temperature].
The Note prohibits boron dilution by introduction of coolant into the RCS with boron concentration less than that required to meet the minimum boron concentration of LCO 3.9.1, or draining operations when SDC forced flow is stopped.
Note 2 allows one SDC loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no draining operation to further reduce RCS water level and that the capability exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the inoperable loop during a time when these tests are safe and possible.
An OPERABLE SDC loop consists of an SDC pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.
Management of gas voids is important to SDC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
Both SDC pumps may be aligned to the Refueling Water Storage Tank to support filling or draining the refueling cavity or for performance of required testing.
APPLICABILITY Two SDC loops are required to be OPERABLE, and one SDC loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the SDC System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System. MODE 6 requirements, with a water level 23 ft above the reactor vessel flange, are covered in LCO 3.9.4, "Shutdown Cooling and Coolant Circulation - High Water Level."
ACTIONS A.1 and A.2 CEOG STS B 3.9.5-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 SDC and Coolant Circulation - Low Water Level B 3.9.5 BASES ACTIONS (continued)
With SDC loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most SDC problems and is reasonable, based on the low probability of the coolant boiling in that time SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one SDC loop is operating and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, this Surveillance demonstrates that the other SDC loop is OPERABLE.
In addition, during operation of the SDC loop with the water level in the vicinity of the reactor vessel nozzles, the SDC loop flow rate determination must also consider the SDC pump suction requirements.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the SDC System in the control room.
Verification that the required loops are OPERABLE and in operation ensures that loops can be placed in operation as needed, to maintain decay heat and retain forced circulation. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is considered reasonable, since other administrative controls are available and have proven to be acceptable by operating experience.
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional SDC pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
REFERENCES 1. FSAR, Section [ ].
- 2. FSAR, Section [ ], ["Gas Management Program."]
CEOG STS B 3.9.5-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, 31 days the piping is filled with water from the pump discharge valve to the injection valve.
SR 3.5.1.21 Verify each ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.1.32 Verify ADS [air supply header] pressure is 31 days
[90] psig.
SR 3.5.1.43 [ Verify the [RHR] System cross tie valve[s] [is] 31 days ]
closed and power is removed from the valve operator[s].
SR 3.5.1.54 [ Verify each LPCI inverter output voltage is 31 days ]
[570] V and [630] V while supplying the respective bus.
SR 3.5.1.65 -------------------------------NOTE------------------------------
Not required to be performed if performed within the previous 31 days.
Verify each recirculation pump discharge valve [and Once each startup bypass valve] cycles through one complete cycle of prior to exceeding full travel [or is de-energized in the closed position]. 25% RTP BWR/4 STS 3.5.1-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.76 Verify the following ECCS pumps develop the [In accordance specified flow rate [against a system head with the Inservice corresponding to the specified reactor pressure]. Testing Program or 92 days]
[System Head No. Corresponding of to a Reactor System Flow Rate Pumps Pressure of]
Core Spray [4250] gpm [1] [113] psig LPCI [17,000] gpm [2] [20] psig SR 3.5.1.87 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [reactor pressure] [1020] and 92 days
[920] psig, the HPCI pump can develop a flow rate
[4250] gpm [against a system head corresponding to reactor pressure].
SR 3.5.1.98 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [reactor pressure] [165] psig, the HPCI [18] months pump can develop a flow rate [4250] gpm [against a system head corresponding to reactor pressure].
BWR/4 STS 3.5.1-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.109 -------------------------------NOTE------------------------------
Vessel injection/spray may be excluded.
Verify each ECCS injection/spray subsystem [18] months actuates on an actual or simulated automatic initiation signal.
SR 3.5.1.1110 -------------------------------NOTE------------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated [18] months automatic initiation signal.
SR 3.5.1.1211 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each ADS valve opens when manually [18] months [on a actuated. STAGGERED TEST BASIS for each valve solenoid]
BWR/4 STS 3.5.1-6 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for each required core spray (CS) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> subsystem, the:
- a. Suppression pool water level is
[12 ft 2 inches] or
- b. -------------------------NOTE------------------------------
Only one required CS subsystem may take credit for this option during OPDRVs.
Condensate storage tank water level is [12 ft].
SR 3.5.2.3 Verify, for each required ECCS injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the injection valve.
SR 3.5.2.43 Verify each required ECCS injection/spray 31 days subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.54 Verify each required ECCS pump develops the [In accordance specified flow rate [against a system head with the Inservice corresponding to the specified reactor pressure]. Testing Program or 92 days]
[System Head No. Corresponding of to a Reactor System Flow Rate Pumps Pressure of]
CS [4250] gpm [1] [113] psig LPCI [7700] gpm [1] [20] psig BWR/4 STS 3.5.2-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.65 -------------------------------NOTE------------------------------
Vessel injection/spray may be excluded.
Verify each required ECCS injection/spray [18] months subsystem actuates on an actual or simulated automatic initiation signal.
BWR/4 STS 3.5.2-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water 31 days from the pump discharge valve to the injection valve.
SR 3.5.3.21 Verify each RCIC System manual, power operated, 31 days and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.3.32 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [reactor pressure] [1020] psig and 92 days
[920] psig, the RCIC pump can develop a flow rate
[400] gpm [against a system head corresponding to reactor pressure].
SR 3.5.3.43 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [reactor pressure] [165] psig, the RCIC [18] months pump can develop a flow rate [400] gpm [against a system head corresponding to reactor pressure].
SR 3.5.3.54 -------------------------------NOTE------------------------------
Vessel injection may be excluded.
Verify the RCIC System actuates on an actual or [18] months simulated automatic initiation signal.
BWR/4 STS 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Hot Shutdown B 3.4.8 BASES LCO (continued) shutdown cooling mode for removal of decay heat. In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy.
Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
Note 1 permits both RHR shutdown cooling subsystems to be removed from operation for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.
APPLICABILITY In MODE 3 with reactor steam dome pressure below [the RHR cut in permissive pressure] (i.e., the actual pressure at which the interlock resets) the RHR System may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant temperature.
Otherwise, a recirculation pump is required to be in operation.
In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to [the RHR cut in permissive pressure], this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS - Operating") do not allow placing the RHR shutdown cooling subsystem into operation.
The requirements for decay heat removal in MODES 4 and 5 are discussed in LCO 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown," LCO 3.9.8, "Residual Heat Removal (RHR) - High Water Level," and LCO 3.9.9, "Residual Heat Removal (RHR) - Low Water Level."
BWR/4 STS B 3.4.8-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Hot Shutdown B 3.4.8 BASES ACTIONS (continued)
B.1, B.2, and B.3 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as permitted by LCO Note 1, reactor coolant circulation by the RHR shutdown cooling subsystem or recirculation pump must be restored without delay.
Until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the coolant circulation function and is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. This will provide assurance of continued temperature monitoring capability.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation. The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period), which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/4 STS B 3.4.8-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Cold Shutdown B 3.4.9 BASES LCO (continued) subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
Note 1 permits both RHR shutdown cooling subsystems to be removed from operation for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.
APPLICABILITY In MODE 4, the RHR Shutdown Cooling System may be operated in the shutdown cooling mode to remove decay heat to maintain coolant temperature below 200°F. Otherwise, a recirculation pump is required to be in operation.
In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS - Operating") do not allow placing the RHR shutdown cooling subsystem into operation.
The requirements for decay heat removal in MODE 3 below the cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.8, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown," LCO 3.9.8, "Residual Heat Removal (RHR) - High Water Level," and LCO 3.9.9, "Residual Heat Removal (RHR) - Low Water Level."
BWR/4 STS B 3.4.9-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Cold Shutdown B 3.4.9 BASES ACTIONS (continued)
B.1 and B.2 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as permitted by LCO Note 1, and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the coolant circulation function and is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. This will provide assurance of continued temperature monitoring capability.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/4 STS B 3.4.9-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES LCO (continued)
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 12).
As noted, LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.
APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS - Shutdown."
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem 2. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 If any one low pressure ECCS injection/spray subsystem is inoperable, or if one LPCI pump in both LPCI subsystems is inoperable, the inoperable subsystem(s) must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time BWR/4 STS B 3.5.1-5 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES ACTIONS (continued) is based on a reliability study (Ref. 1213) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
B.1 and B.2 If the inoperable low pressure ECCS subsystem cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2 If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY immediately is therefore required when HPCI is inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. If the OPERABILITY of the RCIC System cannot be verified, however, Condition G must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on a reliability study cited in Reference 12 13 and has been found to be acceptable through operating experience.
BWR/4 STS B 3.5.1-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES ACTIONS (continued)
D.1 and D.2 If any one low pressure ECCS injection/spray subsystem, or one LPCI pump in both LPCI subsystems, is inoperable in addition to an inoperable HPCI System, the inoperable low pressure ECCS injection/spray subsystem or the HPCI System must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, adequate core cooling is ensured by the OPERABILITY of the ADS and the remaining low pressure ECCS subsystems. However, the overall ECCS reliability is significantly reduced because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since both a high pressure system (HPCI) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the HPCI System or the low pressure ECCS injection/spray subsystem to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 13 and has been found to be acceptable through operating experience.
E.1 The LCO requires seven ADS valves to be OPERABLE in order to provide the ADS function. Reference 13 14 contains the results of an analysis that evaluated the effect of one ADS valve being out of service.
Per this analysis, operation of only six ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study cited in Reference 12 13 and has been found to be acceptable through operating experience.
F.1 and F.2 If any one low pressure ECCS injection/spray subsystem, or one LPCI pump in both LPCI subsystems, is inoperable in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem. However, overall ECCS reliability is reduced because a single active component failure concurrent with a design basis BWR/4 STS B 3.5.1-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES ACTIONS (continued)
LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure system (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the low pressure ECCS subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 13 and has been found to be acceptable through operating experience.
G.1 and G.2 If any Required Action and associated Completion Time of Condition C, D, E, or F is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dome pressure reduced to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
H.1 When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand.
This will also prevent a water hammer following an ECCS initiation signal.
One acceptable method of ensuring that the lines are full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.
BWR/4 STS B 3.5.1-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.21 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would only affect a single subsystem. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.1.32 Verification every 31 days that ADS air supply header pressure is
[90] psig ensures adequate air pressure for reliable ADS operation.
The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 70% of design pressure (Ref. 11). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of
[90] psig is provided by the ADS instrument air supply. The 31 day Frequency takes into consideration administrative controls over operation of the air system and alarms for low air pressure.
BWR/4 STS B 3.5.1-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.43 Verification every 31 days that the RHR System cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.
Acceptable methods of removing power to the operator include de-energizing breaker control power or racking out or removing the breaker.
If the RHR System cross tie valve is open or power has not been removed from the valve operator, both LPCI subsystems must be considered inoperable. The 31 day Frequency has been found acceptable, considering that these valves are under strict administrative controls that will ensure the valves continue to remain closed with either control or motive power removed.
SR 3.5.1.54 Verification every 31 days that each LPCI inverter output has a voltage of
[570] V and [630] V while supplying its respective bus demonstrates that the AC electrical power is available to ensure proper operation of the associated LPCI inboard injection and minimum flow valves and the recirculation pump discharge valve. Each inverter must be OPERABLE for the associated LPCI subsystem to be OPERABLE. The 31 day Frequency has been found acceptable based on engineering judgment and operating experience.
SR 3.5.1.65 Cycling the recirculation pump discharge [and bypass] valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will close when required. Upon initiation of an automatic LPCI subsystem injection signal, these valves are required to be closed to ensure full LPCI subsystem flow injection in the reactor via the recirculation jet pumps. De-energizing the valve in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing the valve include de-energizing breaker control power, racking out the breaker or removing the breaker.
BWR/4 STS B 3.5.1-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The specified Frequency is once during reactor startup before THERMAL POWER is > 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed if the last performance was more than 31 days ago. Therefore, implementation of this Note requires this test to be performed during reactor startup before exceeding 25% RTP. Verification during reactor startup prior to reaching
> 25% RTP is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability of these valves. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.
SR 3.5.1.76, SR 3.5.1.87, and SR 3.5.1.98 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses.
The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 10.
The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established during preoperational testing.
The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system.
Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure must be [920] psig to perform SR 3.5.1.8 7 and [150] psig to perform SR 3.5.1.98. Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the BWR/4 STS B 3.5.1-11 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) time allowed to satisfactorily perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.8 7 and SR 3.5.1.9 8 are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.7 6 and SR 3.5.1.8 7 is in accordance with the Inservice Testing Program requirements. The 18 month Frequency for SR 3.5.1.9 8 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.109 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
BWR/4 STS B 3.5.1-12 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
SR 3.5.1.1110 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.12 11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown.
SR 3.5.1.1211 A manual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or bypass valve or by a change in the measured flow or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to BWR/4 STS B 3.5.1-13 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) perform this SR. Adequate pressure at which this SR is to be performed is [920 psig] (the pressure recommended by the valve manufacturer).
Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Reactor startup is allowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.11 10 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Frequency of 18 months on a STAGGERED TEST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage.
Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. FSAR, Section [6.3.2.2.3].
- 2. FSAR, Section [6.3.2.2.4].
- 3. FSAR, Section [6.3.2.2.1].
- 4. FSAR, Section [6.3.2.2.2].
- 5. FSAR, Section [15.2.8].
- 6. FSAR, Section [15.6.4].
- 7. FSAR, Section [15.6.5].
- 9. FSAR, Section [6.3.3].
BWR/4 STS B 3.5.1-14 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 BASES REFERENCES (continued)
- 10. 10 CFR 50.46.
- 11. FSAR, Section [7.3.1.2.2].
- 12. FSAR, Section [ ], ["Gas Management Program."]
1213. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
1314. FSAR, Section [6.3.3.3].
BWR/4 STS B 3.5.1-15 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the Core Spray (CS) System and the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR)
System is provided in the Bases for LCO 3.5.1, "ECCS - Operating."
APPLICABLE The ECCS performance is evaluated for the entire spectrum of break SAFETY sizes for a postulated loss of coolant accident (LOCA). The long term ANALYSES cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one low pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. It is reasonable to assume, based on engineering judgement, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. To provide redundancy, a minimum of two low pressure ECCS injection/spray subsystems are required to be OPERABLE in MODES 4 and 5.
The low pressure ECCS subsystems satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO Two low pressure ECCS injection/spray subsystems are required to be OPERABLE. The low pressure ECCS injection/spray subsystems consist of two CS subsystems and two LPCI subsystems. Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV. Only a single LPCI pump is required per subsystem because of the larger injection capacity in relation to a CS subsystem. In MODES 4 and 5, the RHR System cross tie valve is not required to be closed.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
As noted, one LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Alignment and operation for decay heat removal BWR/4 STS B 3.5.2-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 REQUIREMENTS The minimum water level of [12 ft 2 inches] required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS System and LPCI subsystem pumps, recirculation volume, and vortex prevention.
With the suppression pool water level less than the required limit, all ECCS injection/spray subsystems are inoperable unless they are aligned to an OPERABLE CST.
When suppression pool level is < [12 ft 2 inches], the CS System is considered OPERABLE only if it can take suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pump. Therefore, a verification that either the suppression pool water level is [12 ft 2 inches] or that CS is aligned to take suction from the CST and the CST contains [150,000] gallons of water, equivalent to 12 ft, ensures that the CS System can supply at least [50,000] gallons of makeup water to the RPV. The CS suction is uncovered at the
[100,000] gallon level. However, as noted, only one required CS subsystem may take credit for the CST option during OPDRVs. During OPDRVs, the volume in the CST may not provide adequate makeup if the RPV were completely drained. Therefore, only one CS subsystem is allowed to use the CST. This ensures the other required ECCS subsystem has adequate makeup volume.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of these SRs was developed considering operating experience related to suppression pool water level and CST water level variations and instrument drift during the applicable MODES.
Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.
SR 3.5.2.3, SR 3.5.2.54, and SR 3.5.2.65 The Bases provided for SR 3.5.1.1, SR 3.5.1.76, and SR 3.5.1.10 9 are applicable to SR 3.5.2.3, SR 3.5.2.54, and SR 3.5.2.65, respectively.
BWR/4 STS B 3.5.2-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.43 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.
REFERENCES 1. FSAR, Section [6.3.2].
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWR/4 STS B 3.5.2-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES BACKGROUND (continued)
The RCIC pump is provided with a minimum flow bypass line, which discharges to the suppression pool. The valve in this line automatically open to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the RCIC System discharge piping is kept full of water. The RCIC System is normally aligned to the CST. The height of water in the CST is sufficient to maintain the piping full of water up to the first isolation valve. The relative height of the feedwater line connection for RCIC is such that the water in the feedwater lines keeps the remaining portion of the RCIC discharge line full of water. Therefore, RCIC does not require a "keep fill" system.
APPLICABLE The function of the RCIC System is to respond to transient events by SAFETY providing makeup coolant to the reactor. The RCIC System is not an ANALYSES Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. The RCIC System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO The OPERABILITY of the RCIC System provides adequate core cooling such that actuation of any of the low pressure ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity for maintaining RPV inventory during an isolation event.
Management of gas voids is important to RCIC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 3).
APPLICABILITY The RCIC System is required to be OPERABLE during MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig, since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure 150 psig, and in MODES 4 and 5, RCIC is not required to be OPERABLE since the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV.
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC System and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
BWR/4 STS B 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES ACTIONS (continued)
A.1 and A.2 If the RCIC System is inoperable during MODE 1, or MODE 2 or 3 with reactor steam dome pressure > [150] psig, and the HPCI System is verified to be OPERABLE, the RCIC System must be restored to OPERABLE status within 14 days. In this Condition, loss of the RCIC System will not affect the overall plant capability to provide makeup inventory at high reactor pressure since the HPCI System is the only high pressure system assumed to function during a loss of coolant accident (LOCA). OPERABILITY of HPCI is therefore verified immediately when the RCIC System is inoperable. This may be performed as an administrative check, by examining logs or other information, to determine if HPCI is out of service for maintenance or other reasons. It does not mean it is necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the HPCI System. If the OPERABILITY of the HPCI System cannot be verified, however, Condition B must be immediately entered. For transients and certain abnormal events with no LOCA, RCIC (as opposed to HPCI) is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of the RPV water level. Therefore, a limited time is allowed to restore the inoperable RCIC to OPERABLE status.
The 14 day Completion Time is based on a reliability study (Ref. 43) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (AOTs). Because of similar functions of HPCI and RCIC, the AOTs (i.e.,
Completion Times) determined for HPCI are also applied to RCIC.
B.1 and B.2 If the RCIC System cannot be restored to OPERABLE status within the associated Completion Time, or if the HPCI System is simultaneously inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dome pressure reduced to [150] psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
BWR/4 STS B 3.5.3-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the RCIC piping, the procedural controls governing system operation, and operating experience.
SR 3.5.3.21 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.3.3 2and SR 3.5.3.43 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated.
The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.
Additionally, adequate steam flow must be passing through the main BWR/4 STS B 3.5.3-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued) turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be [920] psig to perform SR 3.5.3.3 2 and [150] psig to perform SR 3.5.3.43. Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
A 92 day Frequency for SR 3.5.3.3 2 is consistent with the Inservice Testing Program requirements. The 18 month Frequency for SR 3.5.3.4 3 is based on the need to perform the Surveillance under conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.3.54 The RCIC System is required to actuate automatically in order to verify its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
BWR/4 STS B 3.5.3-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued)
The 18 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 33.
- 2. FSAR, Section [5.5.6].
- 3. FSAR, Section [ ], ["Gas Management Program."]
- 43. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
BWR/4 STS B 3.5.3-6 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure.
An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
ACTIONS A.1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B.1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate.
BWR/4 STS B 3.6.2.3-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS (continued)
C.1 and C.2 If the Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.
SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate [7700] gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME Code (Ref. 32). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is [in accordance with the Inservice Testing Program or 92 days].
BWR/4 STS B 3.6.2.3-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES REFERENCES 1. FSAR, Section [6.2].
- 2. FSAR, Section [ ], ["Gas Management Program."]
- 23. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWR/4 STS B 3.6.2.3-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Spray B 3.6.2.4 BASES LCO In the event of a DBA, a minimum of one RHR suppression pool spray subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak pressure below the design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool spray subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool spray subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
Management of gas voids is important to RHR Suppression Pool Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining RHR suppression pool spray subsystems OPERABLE is not required in MODE 4 or 5.
ACTIONS A.1 With one RHR suppression pool spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR suppression pool spray subsystem is adequate to perform the primary containment bypass leakage mitigation function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment bypass mitigation capability. The 7 day Completion Time was chosen in light of the redundant RHR suppression pool spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B.1 With both RHR suppression pool spray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this Condition, there is a substantial loss of the primary containment bypass leakage mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to remove heat from primary containment are available.
BWR/4 STS B 3.6.2.4-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Spray B 3.6.2.4 BASES ACTIONS (continued)
C.1 and C.2 If the inoperable RHR suppression pool spray subsystem cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.2.4.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.
SR 3.6.2.4.2 Verifying each RHR pump develops a flow rate [400] gpm while operating in the suppression pool spray mode with flow through the heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by the ASME Code (Ref. 32). This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice BWR/4 STS B 3.6.2.4-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Spray B 3.6.2.4 BASES SURVEILLANCE REQUIREMENTS (continued) inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is [in accordance with the Inservice Testing Program, but the Frequency must not exceed 92 days].
REFERENCES 1. FSAR, Section [6.2].
- 2. FSAR, Section [ ], ["Gas Management Program."]
- 23. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWR/4 STS B 3.6.2.4-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR - High Water Level B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR) - High Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.
In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal.
APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.
The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE and in operation in MODE 5 with irradiated fuel in the RPV and the water level [23] ft above the RPV flange. Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability.
An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
BWR/4 STS B 3.9.8-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - High Water Level B 3.9.8 BASES SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that the RHR subsystem is in operation and circulating reactor coolant.
The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/4 STS B 3.9.8-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - Low Water Level B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.
APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.
The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 23 ft above the reactor pressure vessel (RPV) flange both RHR shutdown cooling subsystems must be OPERABLE.
An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. To meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to BWR/4 STS B 3.9.9-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - Low Water Level B 3.9.9 BASES ACTIONS (continued) isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
C.1 and C.2 If no RHR subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Completion Time is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.9.9.1 REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/4 STS B 3.9.9-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME H. HPCS and low pressure H.1 Enter LCO 3.0.3. Immediately core spray (LPCS) inoperable.
OR Three or more ECCS injection/spray subsystems inoperable.
OR HPCS System and one or more ADS valves inoperable.
OR Two or more ECCS injection/spray subsystems and one or more ADS valves inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, 31 days the piping is filled with water from the pump discharge valve to the injection valve.
BWR/6 STS 3.5.1-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.21 Verify each ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.1.32 Verify ADS [air receiver] pressure is [150] psig. 31 days SR 3.5.1.43 Verify each ECCS pump develops the specified flow [In accordance rate [against a system head corresponding to the with the Inservice specified reactor pressure]. Testing Program or 92 days]
[System Head Corresponding to a Reactor System Flow Rate Pressure of]
LPCS [7115] gpm [290] psig LPCI [7450] gpm [125] psig HPCS [7115] gpm [445] psig SR 3.5.1.54 -------------------------------NOTE------------------------------
Vessel injection/spray may be excluded.
Verify each ECCS injection/spray subsystem [18] months actuates on an actual or simulated automatic initiation signal.
BWR/6 STS 3.5.1-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.1.65 -------------------------------NOTE------------------------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated [18] months automatic initiation signal.
SR 3.5.1.76 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify each ADS valve opens when manually [18] months on a actuated. STAGGERED TEST BASIS for each valve solenoid BWR/6 STS 3.5.1-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 and D.1 Initiate action to restore Immediately associated Completion [secondary containment] to Time not met. OPERABLE status.
AND D.2 [ Initiate action to restore Immediately ]
one standby gas treatment subsystem to OPERABLE status.
AND D.3 Initiate action to restore Immediately isolation capability in each required [secondary containment] penetration flow path not isolated.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure ECCS 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> injection/spray subsystem, the suppression pool water level is [12.67 ft].
SR 3.5.2.2 Verify, for the required High Pressure Core Spray 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (HPCS) System, the:
- a. Suppression pool water level is [12.67 ft] or
- b. Condensate storage tank water level is [18 ft].
SR 3.5.2.3 Verify, for each required ECCS injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the injection valve.
BWR/6 STS 3.5.2-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.43 Verify each required ECCS injection/spray 31 days subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.2.54 Verify each required ECCS pump develops the [In accordance specified flow rate [against a system head with the Inservice corresponding to the specified reactor pressure]. Testing Program or 92 days]
[System Head Corresponding to a Reactor System Flow Rate Pressure of]
LPCS [7115] gpm [290] psig LPCI [7450] gpm [125] psig HPCS [7115] gpm [445] psig SR 3.5.2.65 -------------------------------NOTE------------------------------
Vessel injection/spray may be excluded.
Verify each required ECCS injection/spray [18] months subsystem actuates on an actual or simulated automatic initiation signal.
BWR/6 STS 3.5.2-3 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water 31 days from the pump discharge valve to the injection valve.
SR 3.5.3.21 Verify each RCIC System manual, power operated, 31 days and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.5.3.32 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [RCIC steam supply pressure] 92 days
[1045] psig and [945] psig, the RCIC pump can develop a flow rate [800] gpm [against a system head corresponding to reactor pressure].
SR 3.5.3.43 -------------------------------NOTE------------------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
Verify, with [RCIC steam supply pressure] [18] months
[165] psig, the RCIC pump can develop a flow rate
[800] gpm [against a system head corresponding to reactor pressure].
SR 3.5.3.54 -------------------------------NOTE------------------------------
Vessel injection may be excluded.
Verify the RCIC System actuates on an actual or [18] months simulated automatic initiation signal.
BWR/6 STS 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Hot Shutdown B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to reduce the temperature of the reactor coolant to 200°F. This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Hot Shutdown condition.
The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal. Each loop consists of a motor driven pump, two heat exchangers in series, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via separate feedwater lines or to the reactor via the LPCI injection path. The RHR heat exchangers transfer heat to the Standby Service Water System (LCO 3.7.1, "[Standby Service Water (SSW)] System and [Ultimate Heat Sink (UHS)]").
APPLICABLE Decay heat removal by the RHR System in the shutdown cooling mode is SAFETY not required for mitigation of any event or accident evaluated in the ANALYSES safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and, when no recirculation pump is in operation, one shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, two heat exchangers in series, and the associated piping and valves. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat.
In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
BWR/6 STS B 3.4.9-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Hot Shutdown B 3.4.9 BASES ACTIONS (continued)
B.1, B.2, and B.3 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as is permitted by LCO Note 1, reactor coolant circulation by the RHR shutdown cooling subsystem or one recirculation pump must be restored without delay.
Until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the coolant circulation function and is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. This will provide assurance of continued temperature monitoring capability.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation. The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period), which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/6 STS B 3.4.9-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Cold Shutdown B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to maintain the temperature of the reactor coolant at 200°F. This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Cold Shutdown condition.
The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal. Each loop consists of a motor driven pump, two heat exchangers in series, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via separate feedwater lines or to the reactor via the LPCI injection path. The RHR heat exchangers transfer heat to the Standby Service Water System.
APPLICABLE Decay heat removal by the RHR System in the shutdown cooling mode is SAFETY not required for mitigation of any event or accident evaluated in the ANALYSES safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and, when no recirculation pump is in operation, one RHR shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, two heat exchangers in series, and the associated piping and valves. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
BWR/6 STS B 3.4.10-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Shutdown Cooling System - Cold Shutdown B 3.4.10 BASES SURVEILLANCE SR 3.4.10.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/6 STS B 3.4.10-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)
- c. Maximum hydrogen generation from zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react,
- d. The core is maintained in a coolable geometry, and
- e. Adequate long term cooling capability is maintained.
The limiting single failures are discussed in Reference 11. For a large break LOCA, failure of ECCS subsystems in Division 1 (LPCS and LPCI-A) or Division 2 (LPCI-B and LPCI-C) due to failure of its associated diesel generator is, in general, the most severe failure. For a small break LOCA, HPCS System failure is the most severe failure. One ADS valve failure is analyzed as a limiting single failure for events requiring ADS operation. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.
The ECCS satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO Each ECCS injection/spray subsystem and eight ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the three LPCI subsystems, the LPCS System, and the HPCS System.
The low pressure ECCS injection/spray subsystems are defined as the LPCS System and the three LPCI subsystems.
With less than the required number of ECCS subsystems OPERABLE during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in 10 CFR 50.46 (Ref. 10) could potentially be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by 10 CFR 50.46 (Ref. 10).
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 12).
As noted, LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the BWR/6 STS B 3.5.1-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, the ADS function is not required when pressure is 150 psig because the low pressure ECCS subsystems (LPCS and LPCI) are capable of providing flow into the RPV below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS - Shutdown."
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCS subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCS subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 If any one low pressure ECCS injection/spray subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced because a single failure in one of the remaining OPERABLE subsystems concurrent with a LOCA may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 1312) that evaluated the impact on ECCS availability by assuming that various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
B.1 and B.2 If the HPCS System is inoperable, and the RCIC System is verified to be OPERABLE (when RCIC is required to be OPERABLE), the HPCS System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with the ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY immediately is therefore required when HPCS is BWR/6 STS B 3.5.1-6 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES ACTIONS (continued) inoperable. This may be performed by an administrative check, by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. However, if the OPERABILITY of the RCIC System cannot be verified and RCIC is required to be OPERABLE, Condition D must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on the results of a reliability study (Ref. 1312) and has been found to be acceptable through operating experience.
C.1 With two ECCS injection subsystems inoperable or one ECCS injection and one ECCS spray subsystem inoperable, at least one ECCS injection/spray subsystem must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced in this Condition because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since the ECCS availability is reduced relative to Condition A, a more restrictive Completion Time is imposed. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on a reliability study, as provided in Reference 1312.
D.1 and D.2 If any Required Action and associated Completion Time of Condition A, B, or C are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
BWR/6 STS B 3.5.1-7 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES ACTIONS (continued)
E.1 The LCO requires eight ADS valves to be OPERABLE to provide the ADS function. Reference 1413 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only seven ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study (Ref. 1312) and has been found to be acceptable through operating experience.
F.1 and F.2 If any one low pressure ECCS injection/spray subsystem is inoperable in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCS and the remaining low pressure ECCS injection/spray subsystems. However, the overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure (ADS) and low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the low pressure ECCS injection/spray subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study (Ref. 1312) and has been found to be acceptable through operating experience.
G.1 and G.2 If any Required Action and associated Completion Time of Condition E or F are not met or if two or more ADS valves are inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dome pressure reduced to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
BWR/6 STS B 3.5.1-8 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES ACTIONS (continued)
H.1 When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCS System, LPCS System, and LPCI subsystems full of water ensures that the systems will perform properly, injecting their full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring the lines are full is to vent at the high points. The 31 day Frequency is based on operating experience, on the procedural controls governing system operation, and on the gradual nature of void buildup in the ECCS piping.
SR 3.5.1.21 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves potentially capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve alignment would only affect a single subsystem. This Frequency has been shown to be acceptable through operating experience.
BWR/6 STS B 3.5.1-9 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.32 Verification every 31 days that ADS air receiver pressure is [150] psig assures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The designed pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 70% of design pressure (Ref. 1514). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of
[150] psig is provided by the ADS Instrument Air Supply System. The 31 day Frequency takes into consideration administrative control over operation of the Instrument Air Supply System and alarms for low air pressure.
SR 3.5.1.43 The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50, Appendix K, criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of 10 CFR 50.46 (Ref. 10).
The pump flow rates are verified against a system head that is equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during LOCAs. These values may be established during pre-operational testing. A 92 day Frequency for this Surveillance is in accordance with the Inservice Testing Program requirements.
SR 3.5.1.54 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance test verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCS, LPCS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup, and actuation of BWR/6 STS B 3.5.1-10 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) all automatic valves to their required positions. This Surveillance also ensures that the HPCS System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
SR 3.5.1.65 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.7 6 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown.
BWR/6 STS B 3.5.1-11 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.76 A manual actuation of each ADS valve is performed to verify that the valve and solenoids are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or bypass valve, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening.
Sufficient time is therefore allowed, after the required pressure and flow are achieved, to perform this test. Adequate pressure at which this test is to be performed is [950] psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Reactor startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. SR 3.5.1.6 5 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Frequency of 18 months on a STAGGERED TEST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform this Surveillance under the conditions that apply just prior to or during a startup from a plant outage.
Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. FSAR, Section [6.3.2.2.3].
- 2. FSAR, Section [6.3.2.2.4].
- 3. FSAR, Section [6.3.2.2.1].
- 4. FSAR, Section [6.3.2.2.2].
- 5. FSAR, Section [15.2.8].
BWR/6 STS B 3.5.1-12 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Operating B 3.5.1 BASES REFERENCES (continued)
- 6. FSAR, Section [15.6.4].
- 7. FSAR, Section [15.6.5].
- 9. FSAR, Section [6.3.3].
- 10. 10 CFR 50.46.
- 11. FSAR, Section [6.3.3.3].
- 12. FSAR, Section [ ], ["Gas Management Program."]
1312. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCO's for ECCS Components,"
December 1, 1975.
1413. FSAR, Section [6.3.3.7.8].
1514. FSAR, Section [7.3.1.1.1.4.2].
BWR/6 STS B 3.5.1-13 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING SYSTEM (RCIC)
B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the High Pressure Core Spray (HPCS) System, Low Pressure Core Spray (LPCS) System, and low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS - Operating."
APPLICABLE The ECCS performance is evaluated for the entire spectrum of break SAFETY sizes for a postulated loss of coolant accident (LOCA). The long term ANALYSES cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one low pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. It is reasonable to assume, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. To provide redundancy, a minimum of two low pressure ECCS injection/spray subsystems are required to be OPERABLE in MODES 4 and 5.
The ECCS satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO Two ECCS injection/spray subsystems are required to be OPERABLE.
The ECCS injection/spray subsystems are defined as the three LPCI subsystems, the LPCS System, and the HPCS System. The LPCS System and each LPCI subsystem consist of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV.
The HPCS System consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV.
Management of gas voids is important to ECCS OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
As noted, one LPCI subsystem (A or B) may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Alignment and operation for decay heat removal includes when the required RHR pump is not operating or when the system is realigned from or to the RHR shutdown cooling mode. This allowance is necessary since the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually BWR/6 STS B 3.5.2-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 ECCS - Shutdown B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
When the suppression pool level is < [12.67 ft], the HPCS System is considered OPERABLE only if it can take suction from the CST and the CST water level is sufficient to provide the required NPSH for the HPCS pump. Therefore, a verification that either the suppression pool water level is [12.67 ft] or the HPCS System is aligned to take suction from the CST and the CST contains [170,000] gallons of water, equivalent to 18 ft, ensures that the HPCS System can supply makeup water to the RPV.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of these SRs was developed considering operating experience related to suppression pool and CST water level variations and instrument drift during the applicable MODES.
Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.
SR 3.5.2.3, SR 3.5.2.54, and SR 3.5.2.65 The Bases provided for SR 3.5.1.1, SR 3.5.1.43, and SR 3.5.1.5 4 are applicable to SR 3.5.2.3, SR 3.5.2.54, and SR 3.5.2.65, respectively.
SR 3.5.2.43 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.
REFERENCES 1. FSAR, Section [6.3.3.4].
- 2. FSAR, Section [ ], ["Gas Management Program."]
BWR/6 STS B 3.5.2-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES APPLICABLE The function of the RCIC System is to respond to transient events by SAFETY providing makeup coolant to the reactor. The RCIC System is not an ANALYSES Engineered Safety Feature System and no credit is taken in the safety analyses for RCIC System operation. The RCIC System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO The OPERABILITY of the RCIC System provides adequate core cooling such that actuation of any of the ECCS subsystems is not required in the event of RPV isolation accompanied by a loss of feedwater flow. The RCIC System has sufficient capacity to maintain RPV inventory during an isolation event.
Management of gas voids is important to RCIC System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 3).
APPLICABILITY The RCIC System is required to be OPERABLE in MODE 1, and MODES 2 and 3 with reactor steam dome pressure > 150 psig since RCIC is the primary non-ECCS water source for core cooling when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure 150 psig, and in MODES 4 and 5, RCIC is not required to be OPERABLE since the ECCS injection/spray subsystems can provide sufficient flow to the vessel.
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC System and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 and A.2 If the RCIC System is inoperable during MODE 1, or MODES 2 or 3 with reactor steam dome pressure > 150 psig, and the HPCS System is verified to be OPERABLE, the RCIC System must be restored to OPERABLE status within 14 days. In this Condition, loss of the RCIC System will not affect the overall plant capability to provide makeup inventory at high RPV pressure since the HPCS System is the only high pressure system assumed to function during a loss of coolant accident (LOCA). OPERABILITY of the HPCS is therefore verified immediately when the RCIC System is inoperable. This may be performed as an administrative check, by examining logs or other information, to determine if the HPCS is out of service for maintenance or other reasons.
Verification does not require performing the Surveillances needed to BWR/6 STS B 3.5.3-2 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES ACTIONS (continued)
Condition B must be immediately entered. For transients and certain abnormal events with no LOCA, RCIC (as opposed to HPCS) is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of RPV water level. Therefore, a limited time is allowed to restore the inoperable RCIC to OPERABLE status.
The 14 day Completion Time is based on a reliability study (Ref. 34) that evaluated the impact on ECCS availability, assuming that various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (AOTs). Because of the similar functions of the HPCS and RCIC, the AOTs (i.e., Completion Times) determined for the HPCS are also applied to RCIC.
B.1 and B.2 If the RCIC System cannot be restored to OPERABLE status within the associated Completion Time, or if the HPCS System is simultaneously inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dome pressure reduced to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the Reactor Coolant System upon demand. This will also prevent a water hammer following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the RCIC piping, the procedural controls governing system operation, and operating experience.
BWR/6 STS B 3.5.3-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.3.21 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path will exist for RCIC operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the RCIC System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least every 92 days.
The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.3.3 2 and SR 3.5.3.43 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated.
The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.
Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be [920] psig to perform SR 3.5.3.3 2 and [150] psig to perform SR 3.5.3.43. Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed BWR/6 STS B 3.5.3-5 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued) that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
A 92 day Frequency for SR 3.5.3.3 2 is consistent with the Inservice Testing Program requirements. The 18 month Frequency for SR 3.5.3.4 3 is based on the need to perform this Surveillance under the conditions that apply just prior to or during startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.3.54 The RCIC System is required to actuate automatically to perform its design function. This Surveillance verifies that with a required system initiation signal (actual or simulated) the automatic initiation logic of RCIC will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions.
This Surveillance test also ensures that the RCIC System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes vessel injection during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
BWR/6 STS B 3.5.3-6 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RCIC System B 3.5.3 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 33.
- 2. FSAR, Section [5.4.6.2].
- 3. FSAR, Section [ ], ["Gas Management Program."]
- 43. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCO's for ECCS Components,"
December 1, 1975.
BWR/6 STS B 3.5.3-7 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR Containment Spray System B 3.6.1.7 BASES LCO In the event of a Design Basis Accident (DBA), a minimum of one RHR containment spray subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak pressure below design limits. To ensure that these requirements are met, two RHR containment spray subsystems must be OPERABLE. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR containment spray subsystem is OPERABLE when the pump, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
Management of gas voids is important to RHR Containment Spray System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining RHR containment spray subsystems OPERABLE is not required in MODE 4 or 5.
ACTIONS A.1 With one RHR containment spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR containment spray subsystem is adequate to perform the primary containment cooling function.
However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time was chosen in light of the redundant RHR containment capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B.1 With two RHR containment spray subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this Condition, there is a substantial loss of the primary containment bypass leakage mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to remove heat from primary containment are available.
BWR/6 STS B 3.6.1.7-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Containment Spray System B 3.6.1.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.1.7.2 Verifying each RHR pump develops a flow rate [5650] gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. It is tested in the pool cooling mode to demonstrate pump OPERABILITY without spraying down equipment in primary containment. Flow is a normal test of centrifugal pump performance required by the ASME Code (Ref. 32). This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. [The Frequency of this SR is in accordance with the Inservice Testing Program or 92 days.]
SR 3.6.1.7.3 This SR verifies that each RHR containment spray subsystem automatic valve actuates to its correct position upon receipt of an actual or simulated automatic actuation signal. Actual spray initiation is not required to meet this SR. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of the safety function. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the
[18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.6.1.7.4 This Surveillance is performed every 10 years to verify that the spray nozzles are not obstructed and that flow will be provided when required.
The 10 year Frequency is adequate to detect degradation in performance due to the passive nozzle design and its normally dry state and has been shown to be acceptable through operating experience.
REFERENCES 1. FSAR, Section [6.2.1.1.5].
- 2. FSAR, Section [ ], ["Gas Management Program."]
- 32. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWR/6 STS B 3.6.1.7-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below the design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE, assuming the worst case single active failure.
An RHR suppression pool cooling subsystem is OPERABLE when the pump, two heat exchangers, and associated piping, valves, instrumentation, and controls are OPERABLE.
Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY. Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 2).
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
ACTIONS A.1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B.1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate.
BWR/6 STS B 3.6.2.3-2 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS (continued)
C.1 and C.2 If the Required Action and required Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves, in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to being locked, sealed, or secured. A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable, since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable, based on operating experience.
SR 3.6.2.3.2 Verifying each RHR pump develops a flow rate [7450] gpm, while operating in the suppression pool cooling mode with flow through the associated heat exchanger at least every 92 days, ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME (Ref. 32). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is
[in accordance with the Inservice Testing Program or 92 days].
BWR/6 STS B 3.6.2.3-3 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR Suppression Pool Cooling B 3.6.2.3 BASES REFERENCES 1. FSAR, Section [6.2].
- 2. FSAR, Section [ ], ["Gas Management Program."]
- 32. ASME Code for Operation and Maintenance of Nuclear Power Plants.
BWR/6 STS B 3.6.2.3-4 Rev. 3.1, 12/01/05
TSTF-523, Rev. 0 RHR - High Water Level B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR) - High Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of one motor driven pump, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via separate feedwater lines or to the upper containment pool via a common single flow distribution sparger or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the Standby Service Water System. The RHR shutdown cooling mode is manually controlled.
In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal.
APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.
The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE in MODE 5 with irradiated fuel in the RPV and the water level [22 ft 8 inches] above the RPV flange. Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability.
An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
BWR/6 STS B 3.9.8-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - High Water Level B 3.9.8 BASES SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that the RHR subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/6 STS B 3.9.8-4 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - Low Water Level B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of one motor driven pump, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via separate feedwater lines, to the upper containment pool via a common single flow distribution sparger, or to the reactor via the low pressure coolant injection path. The RHR heat exchangers transfer heat to the Standby Service Water System. The RHR shutdown cooling mode is manually controlled.
APPLICABLE With the unit in MODE 5, the RHR System is not required to mitigate any SAFETY events or accidents evaluated in the safety analyses. The RHR System ANALYSES is required for removing decay heat to maintain the temperature of the reactor coolant.
The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and with the water level < 22 ft 8 inches above the RPV flange both RHR shutdown cooling subsystems must be OPERABLE.
An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.
Management of gas voids is important to RHR System OPERABILITY.
Gas voids are managed by plant procedures and processes, as described in the Final Safety Analysis Report (Ref. 1).
Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception for the operating subsystem to be removed from operation every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
BWR/6 STS B 3.9.9-1 Rev. 3.0, 03/31/04
TSTF-523, Rev. 0 RHR - Low Water Level B 3.9.9 BASES ACTIONS (continued) administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, a surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The Completion Time is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
SURVEILLANCE SR 3.9.9.1 REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES 1. FSAR, Section [ ], ["Gas Management Program."] None.
BWR/6 STS B 3.9.9-3 Rev. 3.0, 03/31/04