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{{#Wiki_filter:UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 October 21, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon l\Iuclear 4300 Winfield Road Warrenville, IL 60555 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: RISK-INFORIVIED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AI\ID MD6998)  
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 21, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon l\Iuclear 4300 Winfield Road Warrenville, IL 60555
 
==SUBJECT:==
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RISK-INFORIVIED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AI\ID MD6998)


==Dear Mr. Pardee:==
==Dear Mr. Pardee:==
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively.
 
The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession 1\10. ML090350151).
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS)
The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2." A copy of the related Safety Evaluation is also enclosed.
Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession 1\10. ML090350151).
The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket l\Ios. 50-254 and 50-265  
The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2."
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
                                              ~~
Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket l\Ios. 50-254 and 50-265


==Enclosures:==
==Enclosures:==
: 1. Amendment No. 245 to DPR-29 2. Amendment No. 240 to DPR-30 3. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. DPR-29 The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Exelon Generation Company, LLC, et al. (the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-29 is hereby amended to read as follows: 
: 1. Amendment No. 245 to DPR-29
-2Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 245 . are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
: 2. Amendment No. 240 to DPR-30
FOR THE NUCLEAR REGULATORY COMMISSION
: 3. Safety Evaluation cc w/encls: Distribution via Listserv


Campbell.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. DPR-29
Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation  
: 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.      The application for amendment by Exelon Generation Company, LLC, et al.
(the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.      The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.      There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.      The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.      The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-29 is hereby amended to read as follows:
 
                                                -2 B.      Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 245 . are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
: 3.      This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
                                          ~~CfJJJ Stepte~  Campbell. Chief ro~
Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation


==Attachment:==
==Attachment:==
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance:    October 21, 2009


Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance:
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 Renewed License No. DPR-30
October 21, 2009 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, MIDAMERICAN ENERGY DOCKET NO. QUAD CITIES NUCLEAR POWER STATION, UNIT AMENDMENT TO RENEWED FACILITY OPERATING Amendment No. 240 Renewed License No. DPR-30 The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Exelon Generation Company, LLC, et al. (the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the pUblic, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:
: 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 240
A.      The application for amendment by Exelon Generation Company, LLC, et al. (the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.      The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.      There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the pUblic, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.      The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.      The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
* are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:
FOR THE NUCLEAR REGULATORY COMMISSION  
 
/)v Stephen Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation  
                                                -2 B.      Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 240
* are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
: 3.      This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
                                            /)v   l{kj~~~jj ~((
Stephen Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation


==Attachment:==
==Attachment:==
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: October 21, 2009


Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance:
ATTACHMENT TO LICENSE AMENDMENT NOS. 245 AND 240 RENEWED FACILITY OPERATING LICENSES NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by number and contain marginal lines indicating the areas of change.
October 21, 2009 ATTACHMENT TO LICENSE AMENDMENT NOS. 245 AND 240 RENEWED FACILITY OPERATING LICENSES NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by number and contain marginal lines indicating the areas of change. Remove License DPR-29 Page 4 License DPR-30 Page 4 TSs 3.4.3-1 3.5.1-1 3.5.1-2 3.5.1-3 3.5.3-1 3.6.1.6-1 3.6.1.7-2  
Remove License DPR-29                      License DPR-29 Page 4                              Page 4 License DPR-30                      License DPR-30 Page 4                              Page 4 TSs                                TSs 3.4.3-1                            3.4.3-1 3.5.1-1                            3.5.1-1 3.5.1-2                             3.5.1-2 3.5.1-3                            3.5.1-3 3.5.3-1                            3.5.3-1 3.6.1.6-1                          3.6.1.6-1 3.6.1.7-2                          3.6.1.7-2 3.6.1.8-1                          3.6.1.8-1 3.6.2.3-1                           3.6.2.3-1 3.6.2.4-1                          3.6.2.4-1 3.6.4.1-1                          3.6.4.1-1 3.6.4.3-1                          3.6.4.3-1 3.6.4.3-2                          3.6.4.3-2 3.7.1-2                            3.7.1-2 3.7.4-1                            3.7.4-1 3.7.4-2                            3.7.4-2 3.7.5-1                            3.7.5-1 3.7.6-1                            3.7.6-1 3.8.1-5                            3.8.1-5 3.8.4-3                            3.8.4-3 3.8.7-2                            3.8.7-2


3.6.1.8-1 3.6.2.3-1 3.6.2.4-1 3.6.4.1-1 3.6.4.3-1 3.6.4.3-2
                                                -4 B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.245 , are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Oder without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plans', which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.
F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated july 27, 1979 with supplements dated November 5, 1980, and 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Renewed License No. DPR-29 Amendment No. 245


3.7.1-2 3.7.4-1 3.7.4-2 3.7.5-1 3.7.6-1 3.8.1-5 3.8.4-3 3.8.7-2 License DPR-29 Page 4 License DPR-30 Page 4 TSs 3.4.3-1 3.5.1-1 3.5.1-2 3.5.1-3 3.5.3-1 3.6.1.6-1
                                                -4 B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No,- 24Q are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled:
                      "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2,"
submitted by letter dated May 17, 2006.
F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated .July 27 , 1979 with supplements dated 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Renewed License No. DPR-30 Amendment No. 240


3.6.1.7-2 3.6.1.8-1 3.6.2.3-1 3.6.2.4-1 3.6.4.1-1  
Safety and Relief Valves 3.4.3 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.3  Safety and Relief Valves LCO  3.4.3        The safety function of 9 safety valves shall be OPERABLE.
The relief function of 5 relief valves shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTIONS CONDITION                          REQUIRED ACTION                          COMPLETION TIME A. One relief valve        A.1             Restore the relief                    14 days inoperable.                               valve to OPERABLE status.
B. Required Action and      - - - - - - - - - - - - - - NOTE - - - - - - - - -
associated Completion    LCO 3.0.4.a is not applicable Time of Condition A      when entering MODE 3.
not met.                  ----------------------------
B.1             Be in MODE 3.                         12 hours C. Two or more relief      C.1             Be in MODE 3.                         12 hours valves inoperable.
AND OR C.2              Be in MODE 4.                         36 hours One or more safety valves inoperable.
Quad Cities 1 and 2                        3.4.3-1                               Amendment No245/240


3.6.4.3-1 3.6.4.3-2 3.7.1-2 3.7.4-1 3.7.4-2 3.7.5-1 3.7.6-1 3.8.1-5 3.8.4-3 3.8.7-2 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.245 , are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
ECCS-Operating 3.5.1 3.5       EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1         ECCS-Operating LCO     3.5.1                    Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Oder without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
APPLICABILITY:                   MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure ~ 150 psig.
The combined sets of plans', which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated july 27, 1979 with supplements dated November 5, 1980, and 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan. Renewed License No. DPR-29 Amendment No. 245 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No,-24Q are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:
ACTI ONS
The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
The combined set of plans 1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated .July 27 , 1979 with supplements dated 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan. Renewed License No. DPR-30 Amendment No. 240 3.4.3 Safety and Relief Valves 3.4 REACTOR COOLANT SYSTEM 3.4.3 Safety and Relief LCO 3.4.3 The safety function of 9 safety valves shall be The relief function of 5 relief valves shall be OPERABLE.
LCO 3.0.4.b is not applicable to HPCI.
APPLICABILITY:
CONDITION                                             REQU I RED ACTI ON                             COMPLETION TIME A. One Low Pressure                                 A.1              Restore LPCI pump to                        30 days Coolant Injection                                               OPERABLE status.
MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One relief valve inoperable.
(LPCI) pump inoperable.
A.1 Restore the relief valve to OPERABLE status. 14 days B. Required Action and associated Completion Time of Condition A not met. ----------------------LCO 3.0.4.a is not applicable when entering MODE 3.
B. One LPCI subsystem                               B.1            Restore low pressure                        7 days inoperable for reasons                                           ECCS injection/spray other than Condition                                             subsystem to OPERABLE A.                                                               status.
B.1 Be in MODE 3. 12 hours C. Two or more relief valves inoperable.
OR One Core Spray subsystem inoperable.
OR One or more safety valves inoperable.
C. One LPCI pump in each                           C.1             Restore one LPCI pump                        7 days subsystem inoperable.                                            to OPERABLE status.
C.1 Be in MODE 3. AND C.2 Be in MODE 4. 12 hours 36 hours Quad Cities 1 and 2 3.4.3-1 Amendment No245/240 ECCS-Operating 3.5.1 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS-Operating LCO Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
D. Required Action and                             - - - - - - - - - - - - - - NOTE - - - - - - - - - - -
APPLICABILITY:
associated Completion                            LCO 3.0.4.a is not applicable Time of Condition A,                            when entering MODE 3.
MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig. ACTI ONS ----------------------------------
B, or C not met.
NOTE --------------------------------
0.1             Be in MODE 3.                               12 hours (contlnued)
LCO 3.0.4.b is not applicable to HPCI. CONDITION REQU I RED ACTI ON COMPLETION TIME A. One Low Pressure Coolant Injection (LPCI) pump inoperable.
Quad Cities 1 and 2                                                   3.5.1-1                                 Amendment No. 245/240
A.1 Restore LPCI pump OPERABLE status. to 30 days B. One LPCI subsystem inoperable for reasons other than Condition A. OR One Core Spray subsystem inoperable.
 
B.1 Restore low pressure ECCS injection/spray subsystem to OPERABLE status. 7 days C. One LPCI pump in each subsystem inoperable.
ECCS-Operating 3.5.1 ACTIONS CONDITION                      REQU I RED ACTI ON                        COMPLETION TIME E. Two LPCI subsystems   E.l             Restore one LPCI                       72 hours inoperable for reasons                subsystem to OPERABLE other than Condition                  status.
C.1 Restore one to OPERABLE LPCI pump status. 7 days D. Required Action and associated Completion Time of Condition A, B, or C not met. --------------
C.
NOTE -----------
F. Required Action and   F.1            Be in MODE 3.                          12 hours associated Completion Time of Condition E   AND not met.
LCO 3.0.4.a is not applicable when entering MODE 3. 0.1 Be in MODE 3. 12 hours (contlnued)
F.2             Be in MODE 4.                         36 hours G. HPCI System           G.l             Verify by                             Immediately inoperable.                            administrative means RCIC System is OPERABLE.
Quad Cities 1 and 2 3.5.1-1 Amendment No. 245/240 3.5.1 ECCS-Operating ACTIONS E. CONDITION Two LPCI subsystems inoperable for reasons other than Condition C. E.l REQU I RED ACTI ON Restore one LPCI subsystem to OPERABLE status. COMPLETION 72 hours TIME F. Required Action and associated Completion Time of Condition E not met. F.1 AND F.2 Be Be in in MODE MODE 3. 4. 12 36 hours hours G. HPCI System inoperable.
AND G.2            Restore HPCI System                   14 days to OPERABLE status.
G.l AND G.2 Verify by administrative means RCIC System is OPERABLE.
H. One ADS valve         H.l             Restore ADS valve to                   14 days inoperable.                          OPERABLE status.
Restore HPCI System to OPERABLE status. Immediately 14 days H. I. One ADS valve inoperable.
I. Required Action and    - - - - - - - - - - - - - - NOTE - - - - - - - - -
Required Action and associated Completion Time of Condition G or H not met. H.l Restore ADS valve to OPERABLE status. --------------
associated Completion  LCO 3.0.4.a is not applicable Time of Condition G or when entering MODE 3.
NOTE --------LCO 3.0.4.a is not applicable when entering MODE 3. 1.1 Be in MODE 3. 14 12 days hours J. Two or more inoperable.
H not met.
ADS valves J.l Be in MODE 3. 12 hours J.2 Reduce reactor steam dome pressure to S 150 psig. 36 hours (contlnued)
1.1             Be in MODE 3.                         12 hours J. Two or more ADS valves J.l             Be in MODE 3.                         12 hours inoperable.
Quad Cities 1 and 2 3.5.1-2 Amendment No .245/240 3.5.1 ECCS-Operating ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME K. Two or more low pressure ECCS injection/spray subsystems inoperable for reasons other than Condition C or E. K.1 Enter LCO 3.0.3 Immediately HPCI System and one or more ADS valves inoperable.
J.2             Reduce reactor steam                   36 hours dome pressure to S 150 psig.
(contlnued)
Quad Cities 1 and 2                   3.5.1-2                                 Amendment No .245/240
 
ECCS-Operating 3.5.1 ACTIONS CONDITION             REQUIRED ACTION     COMPLETION TIME K. Two or more low       K.1  Enter LCO 3.0.3    Immediately pressure ECCS injection/spray subsystems inoperable for reasons other than Condition C or E.
HPCI System and one or more ADS valves inoperable.
One or more low pressure ECCS injection/spray subsystems inoperable and one or more ADS valves inoperable.
One or more low pressure ECCS injection/spray subsystems inoperable and one or more ADS valves inoperable.
HPCI System inoperable and either one low pressure ECCS injection/spray subsystem is inoperable or Condition Centered.
HPCI System inoperable and either one low pressure ECCS injection/spray subsystem is inoperable or Condition Centered.
Quad Cities 1 and 2 3.5.1-3 Amendment No.245/240 RCIC System 3.5.3 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO The RCIC System shall be OPERABLE.
Quad Cities 1 and 2             3.5.1-3           Amendment No.245/240
APPLICABI MODE 1, MODES 2 and 3 with reactor steam dome pressure>
 
150 psig. ACTI ONS ----------------------------------
RCIC System 3.5.3 3.5      EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3         RCIC System LCO     3.5.3                  The RCIC System shall be OPERABLE.
NOT E--------------------------------
APPLICABI LITY:                  MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.
LCO 3.0.4.b is not applicable to RCIC. CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable.
ACTI ONS
A.1 Verify by administrative means High Pressure Coolant Injection System is OPERABLE.
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOT E- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
AND A.2 Restore RCIC System to OPERABLE status. Immediately 14 days B. Required Action and associated Completion Time not met. ------------
LCO 3.0.4.b is not applicable to RCIC.
NOTE ---------LCO 3.0.4.a is not applicable when entering MODE 3.
CONDITION                                           REQUIRED ACTION                                 COMPLETION TIME A.     RCIC System                                     A.1             Verify by                                   Immediately inoperable.                                                    administrative means High Pressure Coolant Injection System is OPERABLE.
B.1 Be in MODE 3. 12 hours Quad Cities 1 and 3.5.3-1 Amendment No.245/240 Low Set Relief Valves 3.6.1.6 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low Set Relief Valves LCO 3.6.1.6 The low set relief function of two relief valves shall be OPERABLE.
AND A.2             Restore RCIC System                         14 days to OPERABLE status.
APPLICABI LITY: MODES 1, 2, and 3. ACTI ONS A. B. CONDITION One low set relief valve inoperable.
B.     Required Action and                             - - - - - - - - - - - - NOTE - - - - - - - - -
Required Action and associated Completion Time of Condition A not met. REQUIRED ACTION A.1 Restore low set relief valve to OPERABLE status. -------------
associated Completion                          LCO 3.0.4.a is not applicable Time not met.                                  when entering MODE 3.
NOTE ---------LCO 3.0.4.a is not applicable when entering MODE 3.
B.1             Be in MODE 3.                               12 hours Quad Cities 1 and 2                                                  3.5.3-1                                 Amendment No.245/240
B.1 Be in MODE 3. COMPLETION 14 days 12 hours TIME C. Two low set relief valves inoperable.
 
C.1 AND C.2 Be Be in in MODE MODE 3. 4. 12 36 hours hours Quad Cities 1 and 2 3.6.1.6-1 Amendment No. 245/240 Reactor BUilding-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and Associated Completion Time of Condition C not met. --------------
Low Set Relief Valves 3.6.1.6 3.6   CONTAINMENT SYSTEMS 3.6.1.6   Low Set Relief Valves LCO   3.6.1.6     The low set relief function of two relief valves shall be OPERABLE.
NOTE ----------
APPLICABI LITY:   MODES 1, 2, and 3.
LCO 3.0.4.a is not applicable when entering MODE 3.
ACTI ONS CONDITION                        REQUIRED ACTION                          COMPLETION TIME A. One low set relief       A.1             Restore low set                       14 days valve inoperable.                        relief valve to OPERABLE status.
0.1 Be in MODE 3. 12 hours E. Two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening. E.1 Restore all vacuum breakers in one line to OPERABLE status. 1 hour F. Required Action and Associated Completion Time of Conditions A, B or E not met. F.1 Be in MODE 3. AND F.2 Be in MODE 4. 12 hours 36 hours SURVEILLANCE SURVEI LLANCE FREQUENCY SR 3.6.1.7.1 ------------------
B. Required Action and      - - - - - - - - - - - - -NOTE - - - - - - - - -
NOTES -----------------
associated Completion    LCO 3.0.4.a is not applicable Time of Condition A      when entering MODE 3.
not met.                  ----------------------------
B.1             Be in MODE 3.                         12 hours C. Two low set relief       C.1            Be in MODE 3.                          12 hours valves inoperable.
AND C.2             Be in MODE 4.                         36 hours Quad Cities 1 and 2                     3.6.1.6-1                               Amendment No. 245/240
 
Reactor BUilding-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTI ONS CONDITION                                     REQUIRED ACTION                           COMPLETION TIME D. Required Action and                   - - - - - - - - - - - - - - NOTE - - - - - - - - - -
Associated Completion                LCO 3.0.4.a is not applicable Time of Condition C not              when entering MODE 3.
met.                                ----------------------------
0.1             Be in MODE 3.                           12 hours E. Two lines with one or               E.1            Restore all vacuum                      1 hour more reactor building-                             breakers in one line to-suppression chamber                             to OPERABLE status.
vacuum breakers inoperable for opening.
F. Required Action and                 F.1            Be in MODE 3.                          12 hours Associated Completion Time of Conditions A,               AND B or E not met.
F.2             Be in MODE 4.                           36 hours SURVEILLANCE REQUIREMENTS SURVEI LLANCE                                                         FREQUENCY SR   3.6.1.7.1     - - - - - - - - - - - - - - - - - - NOTES - - - - - - - - - - - - - - - - -
: 1. Not required to be met for vacuum breakers that are open during Surveillances.
: 1. Not required to be met for vacuum breakers that are open during Surveillances.
: 2. Not required to be met for vacuum breakers open when performing their intended function.
: 2. Not required to be met for vacuum breakers open when performing their intended function.
Verify each vacuum breaker is closed. 14 days SR 3.6.1.7.2 Perform a functional test of each vacuum breaker. 92 days (continued)
Verify each vacuum breaker is closed.                                         14 days SR   3.6.1.7.2     Perform a functional test of each vacuum                                       92 days breaker.
Quad Cities 1 and 2 3.6.1.7-2 Amendment No. 245/240 Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6 SYSTEMS 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers LCO Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening. Twelve suppression chamber-to-drywell vacuum breakers shall be closed.
(continued)
MODES 1, 2, and 3. ACTI ONS CONDITION REQUIRED ACT ION COMPLETION TIME A. One required suppression to-drywell vacuum breaker inoperable for opening. A.1 Restore one vacuum breaker to OPERABLE status. 72 hours B. Required Action and associated Completion Time of Conditi on A not met. --------------
Quad Cities 1 and 2                                 3.6.1.7-2                                 Amendment No. 245/240
NOTE ----------
 
Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6   CONTAINMENT SYSTEMS 3.6.1.8   Suppression Chamber-to-Drywell Vacuum Breakers LCO   3.6.1.8      Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening.
Twelve suppression chamber-to-drywell vacuum breakers shall be closed.
APPLICABILITY:      MODES 1, 2, and 3.
ACTI ONS CONDITION                         REQUIRED ACT ION                           COMPLETION TIME A. One required             A.1              Restore one vacuum                      72 hours suppression chamber                      breaker to OPERABLE to-drywell vacuum                         status.
breaker inoperable for opening.
B. Required Action and        - - - - - - - - - - - - - - NOTE  - - - - - - - - - -
associated Completion      LCO 3.0.4.a is not applicable Time of Conditi on A      when entering MODE 3.
not met.                  ----------------------------
B.1              Be in MODE 3.                          12 hours C. One suppression          C.1              Close the open vacuum                  4 hours chamber-to-drywell                        breaker.
vacuum breaker not closed.
D. Required Action and      0.1              Be in MODE 3.                          12 hours associated Completion Time of Condition C      AND not met.
0.2              Be in MODE 4.                          36 hours Quad Cities 1 and 2                        3.6.1.8-1                                Amendment No. 245/240
 
RHR Suppression Pool Cooling 3.6.2.3 3.6  CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO  3.6.2.3      Two RHR suppression pool cooling subsystems shall be OPERABLE.
APPLICABILITY:      MODES 1, 2, and 3.
ACTI ONS CONDITION                          REQUIRED ACTION                            COMPLETION TIME A. One RHR suppression      A.1              Restore RHR                            7 days poo 1 cooling subsY$tem                    suppression pool inoperable.                                cooling subsystem to OPERABLE status.
B. Required Action and        - - - - - - - - - - - - - - NOT E- - - - - - - - - -
associated Completion    LCO 3.0.4.a is not applicable Time of Condition A      when entering MODE 3.
not met.                  ----------------------------
B.1              Be in MODE 3.                          12 hours C. Two RHR suppression      C.1              Restore one RHR                        8 hours poo 1 cooling                              suppression pool subsystems inoperable.                    cooling subsystem to OPERABLE status.
D. Required Action and      0.1              Be in MODE 3.                          12 hours associated Completion Time of Condition C      AND not met.
0.2              Be in      I~ODE  4.                  36 hours Quad Cities 1 and 2                        3.6.2.3-1                                Amendment No. 245/240
 
RHR Suppression Pool Spray 3.6.2.4 3.6  CONTAINMENT SYSTEMS 3.6.2.4 Residua 1 Heat Remova 1 (RHR) Suppres s ion Pool Spray LCO  3.6.2.4      Two RHR suppression pool spray subsystems shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTIONS CONDITION                          REQUIRED ACTION                          COMPLETION TIME A. One RHR suppression      A.1              Restore RHR                            7 days pool spray subsystem                      suppression pool inoperable.                              spray subsystem to OPERABLE status.
B. Two RHR suppression      B.1              Restore one RHR                        8 hours pool spray subsystems                      suppression pool inoperable.                                spray subsystem to OPERABLE status.
C. Required Action and      - - - - - - - - - - - - - - NOTE  - - - - - - - - -
associated Completion    LCO 3.0.4.a is not applicable Time not met.            when entering MODE 3.
C.1            Be in MODE 3.                          12 hours Quad Cities 1 and 2                      3.6.2.4-1                                Amendment No. 245/240
 
Secondary Containment 3.6.4.1 3.6  CONTAINMENT SYSTEMS 3.6.4.1  Secondary Containment LCO  3.6.4.1      The secondary containment shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS COND ITION                          REQUI RED ACTION                          COMPLETION TIME A. Secondary containment      A .1            Restore secondary                      4 hours inoperable in MODE 1,                      containment to 2, or 3.                                  OPERABLE status.
B. Required Action and        - - - - - - - - - - - - - - NOTE  - - - - - - - - - -
associated Completion      LCO 3.0.4.a is not applicable Time of Condition A        when entering MODE 3.
not met.
B.1            Be in MODE 3.                          12 hours C. Secondary containment      C.1            - - - - - - - - NOTE  - - - - - - -
inoperable during                          LCO 3.0.3 is not movement of recently                      applicable.
irradiated fuel assemblies in the secondary containment                      Suspend movement of                    Immediately or during OPDRVs.                          recently irradiated fuel assemblies in the secondary containment.
C.2            Initiate action to                      Immediately suspend OPDRVs.
Quad Cities 1 and 2                      3.6.4.1-1                                Amendment No. 245/240
 
SGT System 3.6.4.3 3.6  CONTAINMENT SYSTEMS 3.6.4.3  Standby Gas Treatment (SGT) System LCD  3.6.4.3      Two SGT subsystems shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION                          REQUIRED ACTION                          COMPLETION TIME A. One SGT subsystem          A.1            Restore SGT                            7 days inoperable.                                subsystem to OPERABLE status.
B. Required Action and        - - - - - - - - - - - - - - NOT E- - - - - - - - - -
associated Completion      LCO 3.0.4.a is not applicable Time of Condition A        when entering MODE 3.
not met in MODE 1, 2, or 3.
B.1            Be in MODE 3.                          12 hours C. Required Action and        - - - - - - - - - - - -NOTE  - - - - - - - - - -
associated Completion      LCO 3.0.3 is not applicable.
Time of Condition A not met during movement of recently      C.1            Place OPERABLE SGT                    Immediately i rradi ated fuel                          subsystem in assemblies in the                          operation.
secondary containment or during OPDRVs.
(continued)
Quad Cities 1 and 2                      3.6.4.3-1                                Amendment No. 245/240
 
SGT System 3.6.4.3 ACTIONS CONDITION                    REQUIRED ACTION                          COMPLETION TIME C.  (continued)          C.2.1          Suspend movement of                    Immediately recently irradiated fuel assemblies in secondary containment.
C.2.2          Initiate action to                    Immediately suspend OPDRVs.
D. Two SGT subsystems    D.1            Restore one SGT                        1 hour inoperable in MODE 1,                subsystem to 2, or 3.                              OPERABLE status.
E. Required Action and  - - - - - - - - - - - - - - NOTE  - - - - - - - - -
associated Completion LCO 3.0.4.a is not applicable Time of Condition D  when entering MODE 3.
not met.
E.1            Be in MODE 3.                          12 hours F. Two SGT subsystems    F.1            - - - - - - - -NOTE  - - - - - -
inoperable during                    LCO 3.0.3 is not movement of recently                  applicable.
i rradi ated fuel assemblies in the secondary containment                Suspend movement of                    Immediately or during OPDRVs.                    recently irradiated fuel assemblies in secondary containment.
F.2            Initiate action to                    Immediately suspend OPDRVs.
Quad Cities 1 and 2                  3.6.4.3-2                                Amendment No. 245/240
 
RHRSW System 3.7.1 ACTIONS CONDITION                          REQUIRED ACTION                          COMPLETION TIME D. Required Action and        - - - - - - - - - - - - - - NOT E- - - - - - - - - -
associated Completion    LCO 3.0.4.a is not applicable Time of Conditions A,    when entering MODE 3.
B, or C not met.
D.1              Be in MODE 3.                          12 hours E. Both RHRSW subsystems    E.1             - - - - - - - - NOTE  - - - - - - -
inoperable for reasons                    Enter applicable other than                                Conditions and Condition B.                              Required Actions of LCO 3.4.7 for RHR shutdown cooling subsystems made inoperable by RHRSW System.
Restore one RHRSW                      8 hours subsystem to OPERABLE status.
F. Required Action and        F.1            Be in MODE 3.                          12 hours associated Completion Time of Condition E      AND not met.                                                                          36 hours F.2              Be in MODE 4.
SURVEILLANCE REQUIREMENTS SURVEI LLANCE                                                    FREQUENCY SR  3.7.1.1    Verify each RHRSW manual and power operated                            31 days valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
Quad Cities 1 and 2                        3.7.1-2                                Amendment No. 245/240
 
CREV System 3.7.4 3.7  PLANT SYSTEMS 3.7.4    Control Room Emergency Ventilation (CREV) System LCO  3.7.4        The CREV System shall be OPERABLE.
                    - - - - - - - - - - - - - - NOT E- - - - - - - - - - - - - - - - - -
The main control room envelope (CRE) boundary may be opened intermittently under administrative control.
APPLICABILITY:      MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTI ONS CONDITION                                    REQUI RED ACTION                          COMPLETION TIME A. CREV System inoperable                A.1            Restore CREV System                    7 days in MODE 1, 2, or 3 for                                to OPERABLE status.
reasons other than Condition C.
B. Required Action and                  - - - - - - - - - - - - - - NOT E- - - - - - - - - -
associated Completion                LCO 3.0.4.a is not applicable Time of Condition A                  when entering MODE 3.
not met in MODE 1, 2, or 3.
B.1            Be in MODE 3.                          12 hours C. CREV system                          C.1            Initiate action to                      Immediately inoperable due to                                    implement mitigating inoperable CRE                                        actions.
boundary in MODE 1, 2, or 3.
C.2            Verify mitigating                      24 hours actions ensure CRE occupant exposures to radiological, chemi cal, and smoke hazards wi 11 not exceed 1 i mits (continued)
Quad Cities 1 and 2                                    3.7.4-1                                Amendment No. 245/240
 
CREV System 3.7.4 ACTIONS (continued)
CONDITION                        REQUIRED ACTION                          COMPLETION TIME C.  (continued)              C.3            Restore CRE boundary                  90 days to OPERABLE status D. Required Action and      D.1            Be in MODE 3.                         12 hours associated Completion Time of Condition C not met in MODE 1, 2, or 3.                    0.2            Be in MODE 4.                          36 hours E. CREV System inoperable  - - - - - - - - - - - -NOTE  - - - - - - - - - - - -
during movement of      LCO 3.0.3 is not applicable.
recently irradiated fuel assembl i es in the secondary containment    E.1            Suspend movement of                    Immediately or during OPDRVs.                        recently irradiated fuel assemblies in the secondary containment.
CREV System inoperable due to an inoperable    AND CRE boundary during movement of recently    E.2            Initiate action to                    Immediately irradiated fuel                          suspend OPDRVs.
assemblies in the secondary containment or during OPDRVs.
Quad Cities 1 and 2                      3.7.4-2                                Amendment No. 245/240
 
Control Room Emergency Ventilation AC System 3.7.5 3.7  PLANT SYSTEMS 3.7.5  Control Room Emergency Ventilation Air Conditioning CAC) System LCO  3.7.5        The Control Room Emergency Ventilation AC System shall be OPERABLE.
APPLICABI LITY:    MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel COPDRVs).
ACTIONS CONDITION                          REQUIRED ACTION                            COMPLETION TIME A. Control Room Emergency    A.1            Restore Control Room                    30 days Ventilation AC System                      Emergency Ventilation inoperable in MODE 1,                      AC System to OPERABLE 2, or 3.                                  status.
B. Required Action and       - - - - - - - - - - - - - - NOTE  - - - - - - - - - -
associated Completion     LCO 3.0.4.a is not applicable Time of Condition A       when entering MODE 3.
not met in MODE 1, 2, or 3.
B.1            Be in MODE 3.                          12 hours C. Control Room Emergency    - - - - - - - - - - - - NOT E- - - - - - - - - - - -
Ventilation AC System      LCO 3.0.3 is not applicable.
inoperable during movement of recently irradiated fuel            C.1            Suspend movement of                    Immediately assemblies in the                          recently irradiated secondary containment                      fuel assemblies in or during OPDRVs.                          the secondary containment.
C.2            Initiate action to                      Immediately suspend OPDRVs.
Quad Cities 1 and 2                          3.7.5-1                                Amendment No. 245/240
 
Main Condenser Offgas 3.7.6 3.7  PLANT SYSTEMS 3.7.6    Main Condenser Offgas LCO  3.7.6          The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be
                      ~ 251,100 ~Ci/second after decay of 30 minutes.
APPLICA8I LITY:      l"lODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.
ACTIONS CONDITION                            REQUIRED ACTION                            COMPLETION TIME A. Gross gamma activity          A.1            Restore gross gamma                      72 hours rate of the noble                              activity rate of the gases not withi n                              noble gases to withi n 1 i mi t.                                      1 i mi t.
: 8. Requi red Action and          8.1            Isolate all main                        12 hours associated Completion                          steam lines.
Time not met.
OR 8.2            Isolate SJAE.                            12 hours OR
                                    - - - - - - - - - - - - - - NOTE - - - - - - - - - -
LCO 3.0.4.a is not applicable when entering MODE 3.
LCO 3.0.4.a is not applicable when entering MODE 3.
B.1 Be in MODE 3. 12 hours C. One suppression chamber-to-drywell vacuum breaker not closed. C.1 Close the open vacuum breaker. 4 hours D. Required Action and associated Completion Time of Condition C not met. 0.1 Be in MODE 3. AND 0.2 Be in MODE 4. 12 hours 36 hours Quad Cities 1 and 3.6.1.8-1 Amendment No. 245/240 RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT 3.6.2.3 Residual SYSTEMS Heat Removal (RHR) Suppression Pool Cooling LCO 3.6.2.3 Two RHR suppression pool OPERABLE.
8.3            8e in l"lODE 3 .                         12 hours Quad Cities 1 and 2                              3.7.6-1                                 Amendment No. 245/240
cooling subsystems shall be APPLICABILITY:
 
MODES 1, 2, and 3. ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression poo 1 cooling subsY$tem inoperable.
AC Sources-Operating 3.8.1 ACTIONS cCONDITION                      REQUIRED ACTION                            COMPLETION TIME F. Required Action and   - - - - - - - - - - - - - - NOTE  - - - - - - - - - -
B. Required Action and associated Completion Time of Condition A not met. C. Two RHR suppression poo 1 cooling subsystems inoperable.
associated Completion  LCO 3.0.4.a is not applicable Time of Condition A,  when entering MODE 3.
D. Required Action and associated Completion Time of Condition C not met. A.1 Restore RHR suppression pool cooling subsystem to OPERABLE status. --------------
B, C, 0, or E not met. ----------------------------
NOT E----------
F.1             Be in MODE 3.                           12 hours G. Three or more required G.1            Enter LCO 3.0.3.                         Immediately AC sources inoperable.
LCO 3.0.4.a is not applicable when entering MODE 3.
Quad Cities 1 and 2                     3.8.1-5                                Amendment No. 245/240
B.1 Be in MODE 3. C.1 Restore one RHR suppression pool cooling subsystem to OPERABLE status. 0.1 Be in MODE 3. AND 0.2 Be in
 
: 4. 7 days 12 hours 8 hours 12 hours 36 hours Quad Cities 1 and 2 3.6.2.3-1 Amendment No. 245/240 RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residua 1 Heat Remova 1 (RHR) Suppres s ion Pool LCO 3.6.2.4 Two RHR suppression pool spray subsystems shall be APPLICABILITY:
DC Sources-Operating 3.8.4 ACTIONS CONDITION                      REQUI RED ACTION                           COMPLETION TIME D. Division 1 or 2        D.1             Restore Division lor                    72 hours 125 VDC electrical                    2 125 VDC electrical power subsystem                        power subsystem to inoperable for reasons                OPERABLE status.
MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression pool spray subsystem inoperable.
other than Conditions B or C.                OR D.2            --------NOTE--------
A.1 Restore RHR suppression pool spray subsystem to OPERABLE status. 7 days B. Two RHR suppression pool spray subsystems inoperable.
Only applicable if the opposite unit is not in MODE 1,2, or 3.
B.1 Restore one RHR suppression pool spray subsystem to OPERABLE status. 8 hours C. Required Action and associated Completion Time not met. --------------
Place associated                        72 hours OPERABLE alternate 125 VDC electrical power subsystem in service.
NOTE --------LCO 3.0.4.a is not applicable when entering MODE 3. ----------------C.1 Be in MODE 3. 12 hours Quad Cities 1 and 2 3.6.2.4-1 Amendment No. 245/240 Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.
E. Opposite unit 125 VDC  E.1             Restore the opposite                    7 days electrical power                      unit 125 VDC subsystem inoperable.                  electrical power subsystem to OPERABLE status.
APPLICABILITY:
F. Required Action and   - - - - - - - - - - - - - -NOTE  - - - - - - - - - -
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
associated Completion  LCO 3.0.4.a is not applicable Ti me not met.        when entering MODE 3.
ACTIONS A. B. C. COND ITION Secondary containment inoperable in MODE 1, 2, or 3. Required Action and associated Completion Time of Condition A not met. Secondary containment inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs. A.1 REQUI RED ACTION Restore secondary containment to OPERABLE status. --------------
F.1            Be in MODE 3.                          12 hours Quad Cities 1 and 2                    3.8.4-3                                Amendment No. 245/240
NOTE ----------
 
LCO 3.0.4.a is not applicable when entering MODE 3. B.1 Be in MODE 3. C.1 --------NOTE -------LCO 3.0.3 is not applicable.
Distribution Systems-Operating 3.8.7 ACTIONS CONDITION                      REQUI RED ACTION                          COMPLETION TIME B. One or more DC        B.1             Restore DC electrical                  2 hours electrical power                      power distribution distribution                          subsystems to                          AND subsystems inoperable.                OPERABLE status.
Suspend movement of recently irradiated fuel assemblies in the secondary containment.
16 hours from discovery of failure to meet LCO 3.8.7.a C. One or more required  - - - - - - - - - - - - - NOTE  - - - - - - - - - - -
COMPLETION TIME 4 hours 12 hours Immediately C.2 Initiate action suspend OPDRVs. to Immediately Quad Cities 1 and 2 3.6.4.1-1 Amendment No. 245/240 SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCD 3.6.4.3 Two SGT subsystems shall be OPERABLE.
opposite unit AC or DC Enter applicable Condition electrical power      and Required Actions of distribution          LCO 3.8.1 when Condition C subsystems inoperable. results in the inoperability of a required offsite circuit.
APPLICABILITY:
C.1             Restore required                        7 days opposite unit AC and DC electrical power distribution subsystems to OPERABLE status.
MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
D. Required Action and    - - - - - - - - - - - - - - NOTE  - - - - - - - - - -
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem inoperable.
associated Completion  LCO 3.0.4.a is not applicable Time of Condition A,  when entering MODE 3.
A.1 Restore SGT subsystem to OPERABLE status. 7 days B. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3. --------------
B, or C not met.
NOT E----------
D.1             Be in MODE 3.                           12 hours E. Two or more electrical E.1             Enter LCO 3.0.3.                        Immediately power distribution subsystems inoperable that, in combination, result in a loss of function.
LCO 3.0.4.a is not applicable when entering MODE 3. B.1 Be in MODE 3. 12 hours C. Required Action and associated Completion Time of Condition A not met during movement of recently i rradi ated fuel assemblies in the secondary containment or during OPDRVs. ------------
Quad Cities 1 and 2                     3.8.7-2                                Amendment No. 245/240
NOTE ---------LCO 3.0.3 is not applicable.
 
C.1 Place OPERABLE SGT subsystem in operation.
          ~p.R REGUl
Immediately (continued)
      ~(;:).         "1)
Quad Cities 1 and 2 3.6.4.3-1 Amendment No. 245/240 SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)
UNITED STATES
C.2.1 C.2.2 Suspend movement of recently irradiated fuel assemblies in secondary containment.
  ","",              / 0."....                NUCLEAR REGULATORY COMMISSION f;f                        ('l                        WASHINGTON, D.C. 20555-0001
Initiate action to suspend OPDRVs. Immediately Immediately D. Two SGT subsystems inoperable in MODE 1, 2, or 3. D.1 Restore one SGT subsystem to OPERABLE status. 1 hour E. Required Action and associated Completion Time of Condition D not met. --------------
<l:                          0
NOTE --------LCO 3.0.4.a is not applicable when entering MODE 3. E.1 Be in MODE 3. 12 hours F. Two SGT subsystems inoperable during movement of recently i rradi ated fuel assemblies in the secondary containment or during OPDRVs. F.1 --------NOTE ------LCO 3.0.3 is not applicable.
...                         ~
Suspend movement of recently irradiated fuel assemblies in secondary containment.
III                        ~
Immediately F.2 Initiate action to suspend OPDRVs. Immediately Quad Cities 1 and 2 3.6.4.3-2 Amendment No. 245/240 3.7.1 RHRSW System ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion Time of Conditions A, B, or C not met. --------------
  ~                      p; V"".               ~
NOT E----------
1-?            ~o
LCO 3.0.4.a is not applicable when entering MODE 3. D.1 Be in MODE 3. 12 hours E. Both RHRSW subsystems inoperable for reasons other than Condition B. E.1 --------NOTE -------Enter applicable Conditions and Required Actions of LCO 3.4.7 for RHR shutdown cooling subsystems made inoperable by RHRSW System. Restore one RHRSW subsystem to OPERABLE status. 8 hours F. Required Action and associated Completion Time of Condition E not met. F.1 AND F.2 Be in MODE 3. Be in MODE 4. 12 hours 36 hours SURVEILLANCE SURVEI LLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual and power operated valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
            ****~
31 days Quad Cities 1 and 2 3.7.1-2 Amendment No. 245/240 CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 APPLICABILITY:
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDIVlEI\JT NO.245        TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
ACTI ONS The CREV System shall be OPERABLE. --------------
 
NOT E------------------
==1.0            INTRODUCTION==
The main control room envelope (CRE) boundary may be opened intermittently under administrative control. MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
 
CONDITION REQUI RED ACTION COMPLETION TIME A. CREV System inoperable in MODE 1, 2, or 3 for reasons other than Condition C. A.1 Restore CREV System to OPERABLE status. 7 days B. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3. --------------
By letter dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072830096), as supplemented by letter dated January 30, 2009, (ADAMS Accession No. ML090350151), Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request (LAR) which proposed changes to the technical specifications (TSs) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The January 30, 2009, supplement contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration. The LAR would modify the TSs to risk-informed requirements regarding required action end states. In the request, QCNPS planned to adopt Technical Specification Task Force (TSTF) Change Traveler 423, (Reference (Ref.) 8), to the Boiling-Water Reactor (BWR) Standard Technical Specifications (STS) (NUREG 1433 and NUREG 1434), which was proposed by the TSTF Owners Group on August 12, 2003, on behalf of the industry. TSTF-423 incorporates the BWR Owners Group (BWROG) approved Topical Report (TR) NEDC-32988-A, Revision 2, "Technical Justification to Support Risk Informed Modification to Selected Required Action End States for BWR Plants" (BWROG TR, or Ref. 1), into the BWR STS (NOTE: The changes in TSTF-423 are made with respect to Revision 3 of the BWR STS NUREGs).
NOT E----------
On March 30, 2001 (ADAMS Accession No ML011130309), the Nuclear Regulatory Commission (NRC, the Commission) staff approved the licensee's request to convert the QCNPS TSs to the improved TSs design based on NUREG-1433, Revision 1, "Standard Technical Specifications, General Electric Plants BWR/4," dated April 1995, and on guidance provided in the Commission's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published on July 22,1993 (58 FR 39132). The licensee's October 9,2007, application states that QCNPS TSs are based on NUREG-1433 though it is not identical to the Commission's Policy Statement guidance. Therefore, an adaptation of the referenced document was required.
LCO 3.0.4.a is not applicable when entering MODE 3. B.1 Be in MODE 3. 12 hours C. CREV system inoperable due to inoperable CRE boundary in MODE 1, 2, or 3. C.1 Initiate action to implement mitigating actions. C.2 Verify mitigating actions ensure CRE occupant exposures to radiological, chemi cal, and smoke hazards wi 11 not exceed 1 i mits Immediately 24 hours (continued)
 
Quad Cities 1 and 2 3.7.4-1 Amendment No. 245/240 3.7.4 CREV System ACTIONS (continued)
                                                    -2 TSTF-423 is one of the industry's initiatives developed under the Risk Management Technical Specifications program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and risk management techniques in TSs, while reducing unnecessary burden and making TS requirement~ consistent with the Commission's other risk informed regulatory requirements, in particular, Title 10 of the Code of Federal Regulations (10 CFR), Section 50.65 (Ref. 3), the "Maintenance Rule." Section 50.36(c)(2)(i) of 10 CFR, states, in part: "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow the remedial action permitted by the technical specification until the condition can be met." Plant TSs provide, as part of the remedial action, a completion time (CT) for the plant to either comply with remedial actions or restore compliance with the limiting conditions for operation (LCD). If the LCD or the remedial action cannot be met, then the reactor is required to be shutdown. When the STS and individual plant TSs were written, the shutdown condition, or end state specified, was usually cold shutdown. The BWROG TR provides the technical basis to change certain required end states when the TS Actions for remaining in power operation cannot be met within the CTs. Most of the requested TS changes permit an end state of hot shutdown (Mode 3), if risk is assessed and managed, rather than an end state of cold shutdown (Mode 4), contained in the current TSs. The proposed changes were limited to those end states where: (1) entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical.
CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)
The TSs for QCNPS define five operational modes:
C.3 Restore CRE boundary to OPERABLE status 90 days D. Required Action and associated Completion Time of Condition C not met in MODE 1, 2, or 3. D.1 0.2 Be in MODE 3. Be in MODE 4. 12 hours 36 hours E. CREV System inoperable during movement of recently irradiated fuel assembl i es in the secondary containment or during OPDRVs. CREV System inoperable due to an inoperable CRE boundary during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs. ------------
In general, they are:
NOTE ------------
* Mode 1 - Power Operation. The reactor mode switch is in run position.
LCO 3.0.3 is not applicable.
* Mode 2 - Reactor Startup. The reactor mode switch is in refuel position (with all reactor vessel head closure bolts fully tensioned) or in startup/hot standby position.
E.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
* Mode 3 - Hot Shutdown. The reactor coolant system (RCS) temperature is above 212 degrees F (TS specific) and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tenSioned).
AND E.2 Initiate action to suspend OPDRVs. Immediately Immediately Quad Cities 1 and 2 3.7.4-2 Amendment No. 245/240 Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning CAC) System LCO The Control Room Emergency Ventilation AC System shall be OPERABLE.
* Mode 4 - Cold Shutdown. The RCS temperature is equal to or less than 212 degrees F and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tensioned).
APPLICABI LITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel COPDRVs).
* Mode 5 - Refueling. The reactor mode switch is in shutdown or refuel position, and one or more reactor vessel head closure bolts are less than fully tensioned.
ACTIONS A. B. CONDITION Control Room Emergency Ventilation AC System inoperable in MODE 1, 2, or 3. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, or 3. REQUIRED ACTION A.1 Restore Control Room Emergency Ventilation AC System to OPERABLE status. --------------
Modifying the QCNPS TSs consistent with TSTF-423 allows a Mode 3 end state rather than a Mode 4 end state for selected initiating conditions in order to perform short-duration repairs.
NOTE ----------
Short duration repairs are on the order of 2-to-3 days, but not more than a week.
LCO 3.0.4.a is not applicable when entering MODE 3. B.1 Be in MODE 3. COMPLETION 30 days 12 hours TIME C. Control Room Emergency Ventilation AC System inoperable during movement of recently irradiated fuel assemblies in the secondary containment or during OPDRVs. ------------
The licensee stated in its application that the BWROG TR and TSTF-423, as well as the NRC staff's safety evaluation (SE) (Ref. 6), were applicable to the QCNPS units, and provided justification for incorporation of the proposed changes into the QCNPS Units 1 and 2 TSs.
NOT E------------
 
LCO 3.0.3 is not applicable.
                                                  - 3
C.1 Suspend movement of recently irradiated fuel assemblies in the secondary containment.
 
Immediately C.2 Initiate action suspend OPDRVs. to Immediately Quad Cities 1 and 3.7.5-1 Amendment No. 245/240 Main Condenser Offgas 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Main Condenser Offgas LCO The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be 251,100 after decay of 30 minutes. APPLICA8I LITY: l"lODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.
==2.0      REGULATORY EVALUATION==
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Gross gamma activity rate of the noble gases not withi n 1 i mi t. A.1 Restore gross gamma activity rate of the noble gases to withi n 1 i mi t. 72 hours 8. Requi red Action and associated Completion Time not met. 8.1 Isolate all main steam lines. OR 8.2 Isolate SJAE. OR --------------
 
NOTE ----------
In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following eight specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.
LCO 3.0.4.a is not applicable when entering MODE 3.
The !\IRC staff did not review the LAR with respect to decommissioning, initial notification and written reports, as the licensee did not propose any changes to these specific requirements.
8.3 8e in l"lODE 3. 12 hours 12 hours 12 hours Quad Cities 1 and 3.7.6-1 Amendment No. 245/240 AC Sources-Operating 3.8.1 ACTIONS cCONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and associated Completion Time of Condition A, B, C, 0, or E not met. G. Three or more required AC sources inoperable. --------------
The rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), the LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications. In describing the basis for changing end states, the BWROG TR states:
NOTE ----------
        "Cold shutdown is normally required when an inoperable system or train cannot be restored to an operable status within the allowed time. Going to cold shutdown results in the loss of high pressure core cooling systems, challenges the shutdown heat removal systems, and requires restarting the plant. A more preferred operational mode is one that maintains adequate risk levels while repairs are completed without causing unnecessary challenges to plant equipment during shutdown and startup transitions."
LCO 3.0.4.a is not applicable when entering MODE 3.
In the end state changes under consideration, a problem with a component or train has, or will, result in a failure to meet TSs, and a controlled shutdown is directed because a TS Action statement cannot be met within the TS CT.
F.1 Be in MODE 3. G.1 Enter LCO 3.0.3. 12 hours Immediately Quad Cities 1 and 2 3.8.1-5 Amendment No. 245/240 DC Sources-Operating 3.8.4 ACTIONS CONDITION REQUI RED ACTION COMPLETION TIME D. Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Conditions B or C. D.1 OR Restore Division lor 2 125 VDC electrical power subsystem to OPERABLE status. D.2 Only applicable if the opposite unit is not in MODE 1,2, or 3. 72 hours Place associated OPERABLE alternate 125 VDC electrical power subsystem in service. 72 hours E. Opposite unit 125 VDC electrical power subsystem inoperable.
Most of today's TSs and the design basis analyses were developed under the perception that putting a plant in cold shutdown would result in the safest condition, and the design basis analyses would bound credible shutdown accidents. In the late 1980s and early 1990s, the NRC staff and licensees recognized that this perception was incorrect and took corrective actions to improve shutdown operation. At the same time, the STS were developed and many licensees took action to improve their TSs. Since enactment of a shutdown rule was expected, almost all TS changes involving power operation, including a revised end state requirement, were postponed (Ref. 2). However, in the mid-1990s, the Commission decided a shutdown rule was not necessary in light of industry improvements. Controlling shutdown risk encompasses control of conditions that can cause potential initiating events and responses to those initiating events that do occur. Initiating events are a function of equipment malfunctions and human error.
E.1 Restore the opposite unit 125 VDC electrical power subsystem to OPERABLE status. 7 days F. Required Action and associated Completion Ti me not met. --------------
Responses to events are a function of plant sensitivity, ongoing activities, human error, defense in-depth, and additional equipment malfunctions.
NOTE ----------
In practice, the risk during shutdown operations is often addressed via voluntary actions and application of the Maintenance Rule (Ref. 3). Section 50.65(a)(4) of the Maintenance Rule states, in part:
LCO 3.0.4.a is not applicable when entering MODE 3. F.1 Be in MODE 3. 12 hours Quad Cities 1 and 2 3.8.4-3 Amendment No. 245/240 3.8.7 Distribution Systems-Operating ACTIONS CONDITION REQUI RED ACTION COMPLETION TIME B. One or more DC electrical power distribution subsystems inoperable.
        "Before performing maintenance activities ... , the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of
B.1 Restore DC electrical power distribution subsystems to OPERABLE status. 2 hours AND 16 hours from discovery of failure to meet LCO 3.8.7.a C. One or more required opposite unit AC or DC electrical power distribution subsystems inoperable. -------------
 
NOTE -----------
                                                - 4 the assessment may be limited to structures, systems, and components that a risk informed evaluation process has shown to be significant to public health and safety."
Enter applicable Condition and Required Actions of LCO 3.8.1 when Condition C results in the inoperability of a required offsite circuit. C.1 Restore required opposite unit AC and DC electrical power distribution subsystems to OPERABLE status. 7 days D. Required Action and associated Completion Time of Condition A, B, or C not met. --------------
Regulatory Guide (RG) 1.182 (Ref. 4) provides guidance on implementing the provisions of 10 CFR 50.65(a)(4) by endorsing the revised Section 11 (published separately) to NUMARC 93-01, Revision 2 (Ref.7). The remainder of NUMARC 93-01, Revision 2, was previously endorsed by the NRC staff in RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," March 1997.
NOTE ----------
 
LCO 3.0.4.a is not applicable when entering MODE 3. D.1 Be in MODE 3. 12 hours E. Two or more electrical power distribution subsystems inoperable that, in combination, result in a loss of function.
==3.0     TECHNICAL EVALUATION==
E.1 Enter LCO 3.0.3. Immediately Quad Cities 1 and 2 3.8.7-2 Amendment No. 245/240 REGUl
 
","", / 0.".... UNITED STATES NUCLEAR REGULATORY COMMISSION f;f ('l WASHINGTON, D.C. 20555-0001
The changes proposed in the amendment are consistent with the changes proposed and justified in the BWROG TR and accepted by the !\IRC staff, as documented in the NRC staff's TR SE. The evaluation included in the NRC staffs TR SE, as appropriate and applicable to the changes of TSTF-423, is reiterated here, and differences from the SE are justified. In its application, the licensee commits to the implementation guidance for TSTF-423 contained in TSTF-IG-05-02 (Ref. 9), which addresses a variety of issues, such as considerations and compensatory actions for risk-significant plant configurations. An overview of the generic evaluation and associated risk assessment is provided below, along with a summary of the associated TS changes justified by the BWROG TR.
<l: 0 ... III  p; V"". 1-?
3.1      Risk Assessment The objective of the BWROG TR risk assessment was to show that any risk increases associated with the proposed changes in TS end states are either negligible or negative (Le., a net decrease in risk). The BWROG TR documents a risk-informed analysis of the proposed TS change. Probabilistic Risk Assessment (PRA) results and insights are used, in combination with results of deterministic assessments, to identify and propose changes in "end states" for all BWR plants. This is in accordance with guidance provided in RG 1.174 (Ref. 10) and RG 1.177 (Ref. 5). The three-tiered approach documented in RG 1.177 was followed. The first tier of the three-tiered approach includes the assessment of the risk impact of the proposed change for comparison to acceptance guidelines consistent with the Commission's Safety Goal Policy Statement, as documented in RG 1.174. The first tier aims at ensuring that there are no unacceptable temporary risk increases as a result of the TS change, such as when equipment is taken out of service. The second tier addresses the need to preclude potentially high-risk configurations which could result if equipment is taken out of service concurrently with the equipment out of service as allowed by this TS change. The third tier addresses the application of 10 CFR 50.65 (a)(4) of the Maintenance Rule for identifying risk-significant configurations resulting from maintenance-related activities and taking appropriate compensatory measures to avoid such configurations.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDIVlEI\JT NO.245 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
The proposed TS change invokes a risk assessment because 10 CFR 50.65(a)(4) is applicable to maintenance-related activities and does not cover other operational activities beyond the effect they may have on existing maintenance-related risk.
As discussed in the NRC staff's TR SE, the !\IRC staff found that the BWROG's risk assessment approach used in the BWROG TR was comprehensive and acceptable. In addition, the analyses show that the three-tiered approach criteria for allowing TS changes are met as follows:
 
                                                  -5
* Risk Impact of the Proposed Change (Tier 1): The risk changes associated with the TS changes in TSTF-423, in terms of mean yearly increases in core damage frequency (CDF) and large early release frequency (LERF), are risk neutral or risk beneficial. In addition, there are no significant temporary risk increases, as defined by RG 1.177 criteria, associated with the implementation of the TS end state changes.
* Avoidance of Risk-Significant Configurations (Tier 2): The risk analyses that were performed, which are based on single LCOs, indicate that there are no high-risk configurations associated with the TS end state changes. The reliability of redundant trains is normally covered by a single LCO. When multiple LCOs occur, which affect trains in several systems, the plant's risk-informed configuration risk management program, or the risk assessment and management program implemented in response to 10 CFR 50.65 (a)(4), shall ensure that high-risk configurations are avoided. As part of the implementation of TSTF-423, the licensee has committed to follow Section 11 of NUMARC 93-01, Revision 3, and include guidance in appropriate plant procedures and/or administrative controls to preclude high-risk plant configurations when the plant is at the proposed end state. While the NRC staff has not endorsed Revision 3 to NUMARC 93-01, the NRC staff has endorsed a revised version of NUMARC 93-01, Revision 2, Section 11 in RG 1.182. The !\IRC staff finds that such guidance is adequate for preventing risk-significant plant configurations.
* Configuration Risk Management (Tier 3): The licensee has a commitment in place (as described below), to comply with 10 CFR 50.65 (a)(4) to assess and manage the risk from maintenance activities. This program can support the licensee's decision in selecting the appropriate actions to control risk for most cases in which a risk-informed TS is entered.
The generic risk impact of the end state mode change was evaluated subject to the following assumptions which are incorporated into the TS, TS Bases, and TSTF-IG-05-02:
: a.       The entry into the end state is initiated by the inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification.
: b.      The primary purpose for entering the end state is to correct the initiating condition and return to power as soon as is practical.
: c.       When Mode 3 is entered as the repair end state, the time the reactor coolant pressure is above 500 psig will be minimized. If reactor coolant pressure is above 500 psig for more than 12 hours, the associated plant risk will be assessed and managed.
These assumptions are consistent with typical entries into Mode 3 for short duration repairs, which is the intended use of the TS end state changes. The NRC staff concluded in its TR SE that, in general, going to Mode 3 instead of going to Mode 4 to carry out equipment repairs that are of short duration does not have any adverse effect on plant risk.
In its application, the licensee committed to follow the guidance established in Section 11 of NUMARC 93-01. NUMARC 93-01 provides guidance on implementing the provisions of 10 CFR 50.65(a)(4). The licensee also committed in the January 30,2009, supplement to follow the guidance established in TSTF-IG-05-02. The commitments are restated below:


==1.0 INTRODUCTION==
                                                  -6 COMMITMENT TYPE COMMITMENT                    COI\/IMITTED        ONE TIME          PROGRAMMATIC DATE              ACTION                (YES/NO)
(YES/NO)
EGC will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of            Ongoing              No                    Yes Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 3, July, 2000.
EGC will follow the guidance established in TSTF-IG-05-02, "Implementation Guidance for TSTF 423, Revision 0, 'Technical Specifications End States,NEDC 32988-A,'" Revision 1, March 2007.
Implement              No                    Yes The follOWing statement on Page 2                with no longer applies:                          amendment "If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3."
The NRC staff notes that it has not endorsed NUMARC 93-01, Revision 3, referenced in the first commitment, but has evaluated the language in Section 11 of NUMARC 93-01, Revision 3, and finds that it is consistent with the version of Section 11 that has been endorsed by the NRC staff in RG 1.182. The NRC staff acceptance of this commitment relates only to Section 11 of NUMARC 93-01, Revision 3.
By following the implementation guidance, the licensee will ensure that defense-in-depth is maintained for key safety functions by ensuring availability of Tier 2 systems/equipment necessary for safe shutdown. Therefore, the NRC staff finds the licensee's commitments to be acceptable.
3.2    Request for Additional Information During its review of the application, the staff identified two concerns. First, revising the TS to allow the licensee to remain in Mode 3 indefinitely with inoperable systems would also permit starting up using the allowance of LCO 3.0.4(a) with inoperable systems or equipment. This is inconsistent with the purpose of TSTF-423, which is to allow licensees to remain in Mode 3 (instead of proceeding to Mode 4) while conducting repairs, and then return to Mode 1.


By letter dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072830096), as supplemented by letter dated January 30, 2009, (ADAMS Accession No. ML090350151), Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request (LAR) which proposed changes to the technical specifications (TSs) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The January 30, 2009, supplement contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration.
                                                  - 7 The second concern is that primary containment should not have been treated the same as the other systems included in the TSTF-423. Primary containment was not included in the TSTFs for the pressurized-water reactor designs (Le., TSTF-422 for Combustion Engineering Plants; TSTF-431 for Babcock and Wilcox plants). Unlike the other systems included in TSTF-423, an inoperable primary containment constitutes a loss of one of the three fission product barriers.
The LAR would modify the TSs to risk-informed requirements regarding required action end states. In the request, QCNPS planned to adopt Technical Specification Task Force (TSTF) Change Traveler 423, (Reference (Ref.) 8), to the Boiling-Water Reactor (BWR) Standard Technical Specifications (STS) (NUREG 1433 and NUREG 1434), which was proposed by the TSTF Owners Group on August 12, 2003, on behalf of the industry.
Staying at hot conditions in such an unanalyzed condition is not consistent with maintaining defense-in-depth, which is one of the five key principles of risk-informed regulations in RG 1.174.
TSTF-423 incorporates the BWR Owners Group (BWROG) approved Topical Report (TR) NEDC-32988-A, Revision 2, "Technical Justification to Support Risk Informed Modification to Selected Required Action End States for BWR Plants" (BWROG TR, or Ref. 1), into the BWR STS (NOTE: The changes in TSTF-423 are made with respect to Revision 3 of the BWR STS NUREGs). On March 30, 2001 (ADAMS Accession No ML011130309), the Nuclear Regulatory Commission (NRC, the Commission) staff approved the licensee's request to convert the QCNPS TSs to the improved TSs design based on NUREG-1433, Revision 1, "Standard Technical Specifications, General Electric Plants BWR/4," dated April 1995, and on guidance provided in the Commission's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published on July 22,1993 (58 FR 39132). The licensee's October 9,2007, application states that QCNPS TSs are based on NUREG-1433 though it is not identical to the Commission's Policy Statement guidance.
From the RG perspective, the core damage risks are found to be acceptable; however, the compensatory measures identified (Le., availability of secondary containment, ventilation treatment systems, etc.) do not provide an acceptable defense-in-depth approach and, therefore, an equivalent level of protection, as provided by the primary containment, could not be attained by the compensatory measures.
Therefore, an adaptation of the referenced document was required.
To address these two concerns, the NRC staff issued a request for additional information (RAI)
TSTF-423 is one of the industry's initiatives developed under the Risk Management Technical Specifications program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and risk management techniques in TSs, while reducing unnecessary burden and making TS consistent with the Commission's other informed regulatory requirements, in particular, Title 10 of the Code of Federal Regulations (10 CFR), Section 50.65 (Ref. 3), the "Maintenance Rule." Section 50.36(c)(2)(i) of 10 CFR, states, in part: "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow the remedial action permitted by the technical specification until the condition can be met." Plant TSs provide, as part of the remedial action, a completion time (CT) for the plant to either comply with remedial actions or restore compliance with the limiting conditions for operation (LCD). If the LCD or the remedial action cannot be met, then the reactor is required to be shutdown.
(ADAMS Accession No. ML090080309) to the licensee on January 8, 2009. While the RAI was related to the TSTF-423 review for Clinton Nuclear Power Station, the NRC staff requested that the licensee's response include information for the three facilities with TSTF-423 applications under NRC staff review. The following summarizes the NRC staffs questions, and the licensee's responses for QCNPS:
When the STS and individual plant TSs were written, the shutdown condition, or end state specified, was usually cold shutdown.
RAI1: The licensee was requested to demonstrate how they would prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment.
The BWROG TR provides the technical basis to change certain required end states when the TS Actions for remaining in power operation cannot be met within the CTs. Most of the requested TS changes permit an end state of hot shutdown (Mode 3), if risk is assessed and managed, rather than an end state of cold shutdown (Mode 4), contained in the current TSs. The proposed changes were limited to those end states where: (1) entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical.
The TSs for QCNPS define five operational modes: In general, they are:
* Mode 1 -Power Operation.
The reactor mode switch is in run position.
* Mode 2 -Reactor Startup. The reactor mode switch is in refuel position (with all reactor vessel head closure bolts fully tensioned) or in startup/hot standby position.
* Mode 3 -Hot Shutdown.
The reactor coolant system (RCS) temperature is above 212 degrees F (TS specific) and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tenSioned).
* Mode 4 -Cold The RCS temperature is equal to or less than 212 degrees F and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tensioned).
* Mode 5 Refueling.
The reactor mode switch is in shutdown or refuel position, and one or more reactor vessel head closure bolts are less than fully tensioned.
Modifying the QCNPS TSs consistent with TSTF-423 allows a Mode 3 end state rather than a Mode 4 end state for selected initiating conditions in order to perform short-duration repairs. Short duration repairs are on the order of 2-to-3 days, but not more than a week. The licensee stated in its application that the BWROG TR and TSTF-423, as well as the NRC staff's safety evaluation (SE) (Ref. 6), were applicable to the QCNPS units, and provided justification for incorporation of the proposed changes into the QCNPS Units 1 and 2 TSs. 
-2.0 REGULATORY In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following eight specific categories related to station operation:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. The !\IRC staff did not review the LAR with respect to decommissioning, initial notification and written reports, as the licensee did not propose any changes to these specific requirements.
The rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), the LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
When an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications.
In describing the basis for changing end states, the BWROG TR states: "Cold shutdown is normally required when an inoperable system or train cannot be restored to an operable status within the allowed time. Going to cold shutdown results in the loss of high pressure core cooling systems, challenges the shutdown heat removal systems, and requires restarting the plant. A more preferred operational mode is one that maintains adequate risk levels while repairs are completed without causing unnecessary challenges to plant equipment during shutdown and startup transitions." In the end state changes under consideration, a problem with a component or train has, or will, result in a failure to meet TSs, and a controlled shutdown is directed because a TS Action statement cannot be met within the TS CT. Most of today's TSs and the design basis analyses were developed under the perception that putting a plant in cold shutdown would result in the safest condition, and the design basis analyses would bound credible shutdown accidents.
In the late 1980s and early 1990s, the NRC staff and licensees recognized that this perception was incorrect and took corrective actions to improve shutdown operation.
At the same time, the STS were developed and many licensees took action to improve their TSs. Since enactment of a shutdown rule was expected, almost all TS changes involving power operation, including a revised end state requirement, were postponed (Ref. 2). However, in the mid-1990s, the Commission decided a shutdown rule was not necessary in light of industry improvements.
Controlling shutdown risk encompasses control of conditions that can cause potential initiating events and responses to those initiating events that do occur. Initiating events are a function of equipment malfunctions and human error. Responses to events are a function of plant sensitivity, ongoing activities, human error, in-depth, and additional equipment malfunctions.
In practice, the risk during shutdown operations is often addressed via voluntary actions and application of the Maintenance Rule (Ref. 3). Section 50.65(a)(4) of the Maintenance Rule states, in part: "Before performing maintenance activities
..., the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities.
The scope of 
-the assessment may be limited to structures, systems, and components that a informed evaluation process has shown to be significant to public health and safety." Regulatory Guide (RG) 1.182 (Ref. 4) provides guidance on implementing the provisions of 10 CFR 50.65(a)(4) by endorsing the revised Section 11 (published separately) to NUMARC 93-01, Revision 2 (Ref.7). The remainder of NUMARC 93-01, Revision 2, was previously endorsed by the NRC staff in RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," March 1997. 3.0 TECHNICAL EVALUATION The changes proposed in the amendment are consistent with the changes proposed and justified in the BWROG TR and accepted by the !\IRC staff, as documented in the NRC staff's TR SE. The evaluation included in the NRC staffs TR SE, as appropriate and applicable to the changes of TSTF-423, is reiterated here, and differences from the SE are justified.
In its application, the licensee commits to the implementation guidance for TSTF-423 contained in TSTF-IG-05-02 (Ref. 9), which addresses a variety of issues, such as considerations and compensatory actions for risk-significant plant configurations.
An overview of the generic evaluation and associated risk assessment is provided below, along with a summary of the associated TS changes justified by the BWROG TR. 3.1 Risk Assessment The objective of the BWROG TR risk assessment was to show that any risk increases associated with the proposed changes in TS end states are either negligible or negative (Le., a net decrease in risk). The BWROG TR documents a risk-informed analysis of the proposed TS change. Probabilistic Risk Assessment (PRA) results and insights are used, in combination with results of deterministic assessments, to identify and propose changes in "end states" for all BWR plants. This is in accordance with guidance provided in RG 1.174 (Ref. 10) and RG 1.177 (Ref. 5). The three-tiered approach documented in RG 1.177 was followed.
The first tier of the three-tiered approach includes the assessment of the risk impact of the proposed change for comparison to acceptance guidelines consistent with the Commission's Safety Goal Policy Statement, as documented in RG 1.174. The first tier aims at ensuring that there are no unacceptable temporary risk increases as a result of the TS change, such as when equipment is taken out of service. The second tier addresses the need to preclude potentially high-risk configurations which could result if equipment is taken out of service concurrently with the equipment out of service as allowed by this TS change. The third tier addresses the application of 10 CFR 50.65 (a)(4) of the Maintenance Rule for identifying risk-significant configurations resulting from maintenance-related activities and taking appropriate compensatory measures to avoid such configurations.
The proposed TS change invokes a risk assessment because 10 CFR 50.65(a)(4) is applicable to maintenance-related activities and does not cover other operational activities beyond the effect they may have on existing maintenance-related risk. As discussed in the NRC staff's TR SE, the !\IRC staff found that the BWROG's risk assessment approach used in the BWROG TR was comprehensive and acceptable.
In addition, the analyses show that the three-tiered approach criteria for allowing TS changes are met as follows: 
-5Risk Impact of the Proposed Change (Tier 1): The risk changes associated with the TS changes in TSTF-423, in terms of mean yearly increases in core damage frequency (CDF) and large early release frequency (LERF), are risk neutral or risk beneficial.
In addition, there are no significant temporary risk increases, as defined by RG 1.177 criteria, associated with the implementation of the TS end state changes. Avoidance of Risk-Significant Configurations (Tier 2): The risk analyses that were performed, which are based on single LCOs, indicate that there are no high-risk configurations associated with the TS end state changes. The reliability of redundant trains is normally covered by a single LCO. When multiple LCOs occur, which affect trains in several systems, the plant's risk-informed configuration risk management program, or the risk assessment and management program implemented in response to 10 CFR 50.65 (a)(4), shall ensure that high-risk configurations are avoided. As part of the implementation of TSTF-423, the licensee has committed to follow Section 11 of NUMARC 93-01, Revision 3, and include guidance in appropriate plant procedures and/or administrative controls to preclude high-risk plant configurations when the plant is at the proposed end state. While the NRC staff has not endorsed Revision 3 to NUMARC 93-01, the NRC staff has endorsed a revised version of NUMARC 93-01, Revision 2, Section 11 in RG 1.182. The !\IRC staff finds that such guidance is adequate for preventing risk-significant plant configurations. Configuration Risk Management (Tier 3): The licensee has a commitment in place (as described below), to comply with 10 CFR 50.65 (a)(4) to assess and manage the risk from maintenance activities.
This program can support the licensee's decision in selecting the appropriate actions to control risk for most cases in which a risk-informed TS is entered. The generic risk impact of the end state mode change was evaluated subject to the following assumptions which are incorporated into the TS, TS Bases, and TSTF-IG-05-02: The entry into the end state is initiated by the inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification. The primary purpose for entering the end state is to correct the initiating condition and return to power as soon as is practical. When Mode 3 is entered as the repair end state, the time the reactor coolant pressure is above 500 psig will be minimized.
If reactor coolant pressure is above 500 psig for more than 12 hours, the associated plant risk will be assessed and managed. These assumptions are consistent with typical entries into Mode 3 for short duration repairs, which is the intended use of the TS end state changes. The NRC staff concluded in its TR SE that, in general, going to Mode 3 instead of going to Mode 4 to carry out equipment repairs that are of short duration does not have any adverse effect on plant risk. In its application, the licensee committed to follow the guidance established in Section 11 of NUMARC 93-01. NUMARC 93-01 provides guidance on implementing the provisions of 10 CFR 50.65(a)(4).
The licensee also committed in the January 30,2009, supplement to follow the guidance established in TSTF-IG-05-02.
The commitments are restated below:
COMMITMENT COI\/IMITTED DATE COMMITMENT TYPE ONE TIME ACTION (YES/NO) PROGRAMMATIC (YES/NO) EGC will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 3, July, 2000. Ongoing No Yes EGC will follow the guidance established in TSTF-IG-05-02, "Implementation Guidance for 423, Revision 0, 'Technical Specifications End 32988-A,'" Revision 1, March 2007. The follOWing statement on Page 2 no longer applies: "If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3." Implement with amendment No Yes The NRC staff notes that it has not endorsed NUMARC 93-01, Revision 3, referenced in the first commitment, but has evaluated the language in Section 11 of NUMARC 93-01, Revision 3, and finds that it is consistent with the version of Section 11 that has been endorsed by the NRC staff in RG 1.182. The NRC staff acceptance of this commitment relates only to Section 11 of NUMARC 93-01, Revision 3. By following the implementation guidance, the licensee will ensure that defense-in-depth is maintained for key safety functions by ensuring availability of Tier 2 systems/equipment necessary for safe shutdown.
Therefore, the NRC staff finds the licensee's commitments to be acceptable.
3.2 Request for Additional Information During its review of the application, the staff identified two concerns.
First, revising the TS to allow the licensee to remain in Mode 3 indefinitely with inoperable systems would also permit starting up using the allowance of LCO 3.0.4(a) with inoperable systems or equipment.
This is inconsistent with the purpose of TSTF-423, which is to allow licensees to remain in Mode 3 (instead of proceeding to Mode 4) while conducting repairs, and then return to Mode 1. 
-The second concern is that primary containment should not have been treated the same as the other systems included in the TSTF-423.
Primary containment was not included in the TSTFs for the pressurized-water reactor designs (Le., TSTF-422 for Combustion Engineering Plants; TSTF-431 for Babcock and Wilcox plants). Unlike the other systems included in TSTF-423, an inoperable primary containment constitutes a loss of one of the three fission product barriers.
Staying at hot conditions in such an unanalyzed condition is not consistent with maintaining defense-in-depth, which is one of the five key principles of risk-informed regulations in RG 1.174. From the RG perspective, the core damage risks are found to be acceptable; however, the compensatory measures identified (Le., availability of secondary containment, ventilation treatment systems, etc.) do not provide an acceptable defense-in-depth approach and, therefore, an equivalent level of protection, as provided by the primary containment, could not be attained by the compensatory measures.
To address these two concerns, the NRC staff issued a request for additional information (RAI) (ADAMS Accession No. ML090080309) to the licensee on January 8, 2009. While the RAI was related to the TSTF-423 review for Clinton Nuclear Power Station, the NRC staff requested that the licensee's response include information for the three facilities with TSTF-423 applications under NRC staff review. The following summarizes the NRC staffs questions, and the licensee's responses for QCNPS: RAI1: The licensee was requested to demonstrate how they would prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment.
RAI 2: The licensee was requested to demonstrate how they would maintain an equivalent level of protection while operating in Mode 3 with an inoperable primary containment.
RAI 2: The licensee was requested to demonstrate how they would maintain an equivalent level of protection while operating in Mode 3 with an inoperable primary containment.
On January 30, 2009, the licensee provided their response (ADAMS Accession No. ML090350151) to the NRC staff's RAI as follows: Response to RA11: To prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment, EGC proposed the insertion of the following Note into those Required Actions affected by TSTF-423:
On January 30, 2009, the licensee provided their response (ADAMS Accession No. ML090350151) to the NRC staff's RAI as follows:
NOTE LCO 3.0.4.a is not applicable when entering MODE 3. In addition, the licensee indicated that some of the previously submitted TS pages have been amended since its request to adopt TSTF-423 at QCNPS. Accordingly, the licensee provided revised versions of the TS pages that included the original TSTF-423 adoption markups and the above Note. Response to RAI 2: The licensee indicated that it had evaluated its requests to amend station TS for primary containment and decided to withdraw its request to amend this TS. Because Mode 3 is no longer the requested end state for primary containment, the licensee determined that it is necessary to revise its original commitment to follow guidance established in TSTF-IG-05-02, to indicate that the following statement on Page 2 no longer applies: "If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3."
Response to RA11: To prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment, EGC proposed the insertion of the following Note into those Required Actions affected by TSTF-423:
-Conclusion:
NOTE LCO 3.0.4.a is not applicable when entering MODE 3.
In addition, the licensee indicated that some of the previously submitted TS pages have been amended since its request to adopt TSTF-423 at QCNPS. Accordingly, the licensee provided revised versions of the TS pages that included the original TSTF-423 adoption markups and the above Note.
Response to RAI 2: The licensee indicated that it had evaluated its requests to amend station TS for primary containment and decided to withdraw its request to amend this TS. Because Mode 3 is no longer the requested end state for primary containment, the licensee determined that it is necessary to revise its original commitment to follow guidance established in TSTF-IG-05-02, to indicate that the following statement on Page 2 no longer applies:
          "If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3."
 
                                                - 8
 
==
Conclusion:==
 
The NRC staff reviewed the licensee's response to the staffs RAls, and found them to be acceptable since the amended station TSs prevent: a) operation in Mode 3 without primary containment, and b) starting up with inoperable systems or equipment.
The NRC staff reviewed the licensee's response to the staffs RAls, and found them to be acceptable since the amended station TSs prevent: a) operation in Mode 3 without primary containment, and b) starting up with inoperable systems or equipment.
3.3 Assessment of TS Changes The following sections discuss the specific changes, and include a synopsis of the STS LCOs. The NRC staff discusses the acceptability of the proposed changes in Section 3.4. 3.3.1 LCO 3.4.3: Safety and Relief Valves The function of the safety valves is to protect the plant against severe overpressurization events. The function of the relief valves is to control RCS pressure during transient conditions to prevent the need for safety valve actuation (except the S/RV: one of the safety valves also functions in relief mode) following such a transient.
3.3     Assessment of TS Changes The following sections discuss the specific changes, and include a synopsis of the STS LCOs.
The NRC staff discusses the acceptability of the proposed changes in Section 3.4.
3.3.1   LCO 3.4.3: Safety and Relief Valves The function of the safety valves is to protect the plant against severe overpressurization events.
The function of the relief valves is to control RCS pressure during transient conditions to prevent the need for safety valve actuation (except the S/RV: one of the safety valves also functions in relief mode) following such a transient.
LCO: The safety function of 9 safety valves and the relief function of 5 relief valves shall be OPERABLE.
LCO: The safety function of 9 safety valves and the relief function of 5 relief valves shall be OPERABLE.
Condition Requiring Entry into End State:
Condition Requiring Entry into End State: If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the relief valve cannot be returned to operable status within that time, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours.
If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the relief valve cannot be returned to operable status within that time, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours. Modification for End State Required Actions: If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the valve cannot be returned to operable status within 14 days, the plant must be placed in Mode 3 within 12 hours. A Note is added to the TS Required Action for B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. If two or more relief valves or one or more safety valves become inoperable, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours. Assessment:
Modification for End State Required Actions: If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the valve cannot be returned to operable status within 14 days, the plant must be placed in Mode 3 within 12 hours. A Note is added to the TS Required Action for B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. If two or more relief valves or one or more safety valves become inoperable, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours.
In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed.
Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. The NRC staff reviewed the PRA evaluation and concluded in its TR SE that the core damage risks are approximately the same or lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 for one inoperable relief valve would cause loss of the High Pressure Coolant Injection (HPCI) system, Reactor Core Isolation Cooling (RCIC) system, and the power conversion systems due to the plant cooldown, and would require activating the Residual Heat Removal (RHR) system.
The NRC staff reviewed the PRA evaluation and concluded in its TR SE that the core damage risks are approximately the same or lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 for one inoperable relief valve would cause loss of the High Pressure Coolant Injection (HPCI) system, Reactor Core Isolation Cooling (RCIC) system, and the power conversion systems due to the plant cooldown, and would require activating the Residual Heat Removal (RHR) system. With one relief valve inoperable, the remaining valves are adequate to perform the required function.
With one relief valve inoperable, the remaining valves are adequate to perform the required function. By remaining in Mode 3, HPCl j RCIC and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared to transitioning to Mode 4. In addition, the plant Emergency Operating Procedures (EOPs) direct the operators to take control of the depressurization function if low pressure injection/spray systems are needed for reactor pressure vessel (RPV) water makeup and cooling. The NRC staff concluded in its TR SE that the change allows repairs of the inoperable relief valve to be performed in a plant operating mode with lower risks.
By remaining in Mode 3, HPCl j RCIC and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared to transitioning to Mode 4. In addition, the plant Emergency Operating Procedures (EOPs) direct the operators to take control of the depressurization function if low pressure injection/spray systems are needed for reactor pressure vessel (RPV) water makeup and cooling. The NRC staff concluded in its TR SE that the change allows repairs of the inoperable relief valve to be performed in a plant operating mode with lower risks.
 
-The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in the TS Required Action for B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with this system inoperable.
                                                  - 9 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in the TS Required Action for B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with this system inoperable.
3.3.2 LCO 3.5.1: Emergency Core Cooling Systems (ECCS) -Operating The ECCS systems provide cooling water to the core in the event of a loss-of-coolant accident (LOCA). This set of ECCS TSs provides the operability requirements for the various ECCS subsystems as described below. This TS change would delete the secondary actions. The plant can remain in Mode 3 until the required repair actions are completed.
3.3.2   LCO 3.5.1: Emergency Core Cooling Systems (ECCS) - Operating The ECCS systems provide cooling water to the core in the event of a loss-of-coolant accident (LOCA). This set of ECCS TSs provides the operability requirements for the various ECCS subsystems as described below. This TS change would delete the secondary actions. The plant can remain in Mode 3 until the required repair actions are completed. The reactor is not depressurized.
The reactor is not depressurized.
LCO: Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
LCO: Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
Condition Requiring Entry into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions:
Condition Requiring Entry into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions:
: a. If one Low-Pressure Coolant Injection (LPCI) pump is inoperable, the LPCI pump must be restored to operable status in 30 days (Condition A). b. If one LPCI subsystem is inoperable for reasons other than Condition A or one Core Spray subsystem is inoperable, the low pressure ECCS injection/spray subsystem must be restored to operable status within 7 days (Condition B). c. If one LPCI pump in each subsystem is inoperable, one LPCI pump must be restored to operable status within 7 days (Condition C). d. If two LPCI subsystems are inoperable for reasons other than Condition C, one LPCI subsystem must be restored to operable status within 72 hours (Condition D). e. If the Required Action and associated Completion Time of Condition A, B, C, or D is not met, then place the plant in Mode 3 within 12 hours and in Mode 4 within 36 hours (Condition E). f. If the HPCI system is inoperable, verify immediately by administrative means that the RCIC system is operable and restore the HPCI system to operable status within 14 days (Condition F). g. If one ADS valve is inoperable, it must be restored to operable status within 14 days (Condition G). If the Required Action and associated Completion Time of Condition F or G is not met or two or more ADS valves become inoperable, the plant must be placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours (Condition H). If two or more low pressure ECCS injection/spray subsystems are inoperable for reasons other than Condition C or D; or the HPCI system and one or more ADS
: a.         If one Low-Pressure Coolant Injection (LPCI) pump is inoperable, the LPCI pump must be restored to operable status in 30 days (Condition A).
-10 valves are inoperable; or one or more low pressure ECCS injection/spray subsystems are inoperable and one or more ADS valve are inoperable; or the HPCI system is inoperable and either one low pressure ECCS injection/spray subsystems is inoperable, or Condition C is entered, then LCO 3.0.3 must be entered immediately (Condition I). Modification for End State Required Actions: No change in Required Action for Condition A. No change in Required Action for Condition B. No change in Required Action for Condition C. New Condition 0 states that, if the Required Action and associated Completion Time of Condition A, B, or C are not met, then place the plant in Mode 3 within 12 hours (new Required Action 0.1). The plant is not taken into Mode 4 (cold shutdown).
: b.         If one LPCI subsystem is inoperable for reasons other than Condition A or one Core Spray subsystem is inoperable, the low pressure ECCS injection/spray subsystem must be restored to operable status within 7 days (Condition B).
A Note is added to the TS Required Action 0.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Old Conditions 0, E, F and G are renumbered to E, F, G, and H with no changes to the Required Actions for these Conditions. New Condition I states that, if the Required Action and associated Completion Time of renumbered Conditions G or H are not met, then the plant must be placed in Mode 3 within 12 hours (new Required Action I). The reactor is not depressurized and not taken to Mode 4. A Note is added to the TS Required Action 1.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Old Conditions H and I are renumbered to J and K. The renumbered Condition J states that if two or more ADS valves are inoperable, the plant must be placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours. In renumbered Condition K, Condition 0 is changed to Condition E, in accordance with the reformatting of the "Conditions" column, with no changes to the Required Action and Completion Time for the Condition.
: c.         If one LPCI pump in each subsystem is inoperable, one LPCI pump must be restored to operable status within 7 days (Condition C).
Assessment:
: d.         If two LPCI subsystems are inoperable for reasons other than Condition C, one LPCI subsystem must be restored to operable status within 72 hours (Condition D).
The BWROG TR discusses a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state. The NRC staff's conclusion on the BWROG TR's PRA evaluation described in the NRC staffs TR SE indicates that the core damage risks are lower in Mode 3 than in the current end state of Mode 4. For QCNPS, going to Mode 4 for one ECCS subsystem or one ADS valve would cause loss of the HPCI, RCIC and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system. In addition, plant EOPs direct the operator to take control of the depressurization function if low-pressure injection/spray systems are needed for RPV water makeup and cooling. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions 0.1 and 1.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
: e.         If the Required Action and associated Completion Time of Condition A, B, C, or D is not met, then place the plant in Mode 3 within 12 hours and in Mode 4 within 36 hours (Condition E).
-11 3.3.3 LCD 3.5.3: Reactor Core Isolation Cooling System The function of the RCIC system is to provide reactor coolant makeup during loss of feedwater and other transient events. This TS provides the operability requirements for the RCIC system as described below. In the event of an inoperable RCIC system, the TS change allows the plant to remain in Mode 3 until the repairs are completed.
: f.         If the HPCI system is inoperable, verify immediately by administrative means that the RCIC system is operable and restore the HPCI system to operable status within 14 days (Condition F).
LCD: The RCIC system must be operable during Mode 1. During Modes 2 and 3, RCIC system must be operable when the reactor steam dome pressure is greater than 150 psig. Condition Requiring Entry into End State: If the LCD cannot be met, the following actions must be taken: (a) verify immediately by administrative means that the HPCI system is operable (Required Action A.1), and (b) restore the RCIC system to operable status within 14 days (Required Action A.2). If either or both actions cannot be completed within the allotted time, the plant must be in placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours (Required Actions B.1 and B.2). Modification for End State Required Actions: This TS change keeps the plant in Mode 3 until the required repairs are completed.
: g.         If one ADS valve is inoperable, it must be restored to operable status within 14 days (Condition G).
A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. The reactor steam dome pressure is not reduced to less than 150 psig (delete Required Action B.2). Assessment:
: h.        If the Required Action and associated Completion Time of Condition F or G is not met or two or more ADS valves become inoperable, the plant must be placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours (Condition H).
This change would allow the inoperable RCIC system to be repaired in a plant operating mode with lower risk and without challenging the normal shutdown systems. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed.
: i.        If two or more low pressure ECCS injection/spray subsystems are inoperable for reasons other than Condition C or D; or the HPCI system and one or more ADS
In the NRC staff's TR SE, it concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 3 with reactor steam dome pressure less than 150 psig would cause loss of HPCI and a loss of the power conversion systems, and would require activating the RHR system. By remaining in Mode 3 above 150 psig steam dome pressure, HPCI and the power conversion systems remain available for coolant inventory control and decay heat removal. In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling. The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
 
3.3.4 LCD 3.6.1.6: Low Set Relief Valves The function of low set relief valves is to prevent excessive short-duration relief valve cycling during an overpressure event. The TS provides operability requirements for the two low set relief valves as described below. The TS change allows the plant to remain in Mode 3 until the repairs are completed.
                                                - 10 valves are inoperable; or one or more low pressure ECCS injection/spray subsystems are inoperable and one or more ADS valve are inoperable; or the HPCI system is inoperable and either one low pressure ECCS injection/spray subsystems is inoperable, or Condition C is entered, then LCO 3.0.3 must be entered immediately (Condition I).
Modification for End State Required Actions:
: a.      No change in Required Action for Condition A.
: b.      No change in Required Action for Condition B.
: c.      No change in Required Action for Condition C.
: d.      New Condition 0 states that, if the Required Action and associated Completion Time of Condition A, B, or C are not met, then place the plant in Mode 3 within 12 hours (new Required Action 0.1). The plant is not taken into Mode 4 (cold shutdown). A Note is added to the TS Required Action 0.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Old Conditions 0, E, F and G are renumbered to E, F, G, and H with no changes to the Required Actions for these Conditions.
: e.      New Condition I states that, if the Required Action and associated Completion Time of renumbered Conditions G or H are not met, then the plant must be placed in Mode 3 within 12 hours (new Required Action I). The reactor is not depressurized and not taken to Mode 4. A Note is added to the TS Required Action 1.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
: f.      Old Conditions H and I are renumbered to J and K. The renumbered Condition J states that if two or more ADS valves are inoperable, the plant must be placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours. In renumbered Condition K, Condition 0 is changed to Condition E, in accordance with the reformatting of the "Conditions" column, with no changes to the Required Action and Completion Time for the Condition.
Assessment: The BWROG TR discusses a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state. The NRC staff's conclusion on the BWROG TR's PRA evaluation described in the NRC staffs TR SE indicates that the core damage risks are lower in Mode 3 than in the current end state of Mode 4. For QCNPS, going to Mode 4 for one ECCS subsystem or one ADS valve would cause loss of the HPCI, RCIC and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system. In addition, plant EOPs direct the operator to take control of the depressurization function if low-pressure injection/spray systems are needed for RPV water makeup and cooling.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions 0.1 and 1.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
 
                                                - 11 3.3.3   LCD 3.5.3: Reactor Core Isolation Cooling System The function of the RCIC system is to provide reactor coolant makeup during loss of feedwater and other transient events. This TS provides the operability requirements for the RCIC system as described below. In the event of an inoperable RCIC system, the TS change allows the plant to remain in Mode 3 until the repairs are completed.
LCD: The RCIC system must be operable during Mode 1. During Modes 2 and 3, RCIC system must be operable when the reactor steam dome pressure is greater than 150 psig.
Condition Requiring Entry into End State: If the LCD cannot be met, the following actions must be taken: (a) verify immediately by administrative means that the HPCI system is operable (Required Action A.1), and (b) restore the RCIC system to operable status within 14 days (Required Action A.2). If either or both actions cannot be completed within the allotted time, the plant must be in placed in Mode 3 within 12 hours and the reactor steam dome pressure reduced to less than 150 psig within 36 hours (Required Actions B.1 and B.2).
Modification for End State Required Actions: This TS change keeps the plant in Mode 3 until the required repairs are completed. A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. The reactor steam dome pressure is not reduced to less than 150 psig (delete Required Action B.2).
Assessment: This change would allow the inoperable RCIC system to be repaired in a plant operating mode with lower risk and without challenging the normal shutdown systems. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. In the NRC staff's TR SE, it concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 3 with reactor steam dome pressure less than 150 psig would cause loss of HPCI and a loss of the power conversion systems, and would require activating the RHR system. By remaining in Mode 3 above 150 psig steam dome pressure, HPCI and the power conversion systems remain available for coolant inventory control and decay heat removal. In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling.
The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.4   LCD 3.6.1.6: Low Set Relief Valves The function of low set relief valves is to prevent excessive short-duration relief valve cycling during an overpressure event. The TS provides operability requirements for the two low set relief valves as described below. The TS change allows the plant to remain in Mode 3 until the repairs are completed.
LCD: The low set relief function of two relief valves shall be OPERABLE.
LCD: The low set relief function of two relief valves shall be OPERABLE.
Condition Requiring Entry into End State: If one low set relief valve is inoperable, it must be returned to operability within 14 days. If the low set relief valve cannot be returned to operable
Condition Requiring Entry into End State: If one low set relief valve is inoperable, it must be returned to operability within 14 days. If the low set relief valve cannot be returned to operable
-12status within the allotted time, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours. Modification for End State Required Actions: The TS change would allow the plant to stay in Mode 3 until the required repair actions are completed.
 
The plant would not be taken into Mode 4 (cold shutdown) (delete Required Action B.2). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. (!'Jote: the Required Action for two low set relief valves inoperable was changed from Condition B to new Condition C without changing the Required Action end state.) Assessment:
                                                  - 12 status within the allotted time, the plant must be placed in Mode 3 within 12 hours and in Mode 4 within 36 hours.
In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state was performed.
Modification for End State Required Actions: The TS change would allow the plant to stay in Mode 3 until the required repair actions are completed. The plant would not be taken into Mode 4 (cold shutdown) (delete Required Action B.2). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. (!'Jote: the Required Action for two low set relief valves inoperable was changed from Condition B to new Condition C without changing the Required Action end state.)
In the NRC staff's TR SE, the staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause loss of the HPCI, RCIC, and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system. With one low set relief valve inoperable, the remaining valves are adequate to perform the required function.
Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state was performed. In the NRC staff's TR SE, the staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause loss of the HPCI, RCIC, and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system.
In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling. The NRC staff concluded in its TR SE that this change allows repairs of the inoperable low set relief valve to be performed in a plant operating mode with lower risks. The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.
With one low set relief valve inoperable, the remaining valves are adequate to perform the required function. In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling. The NRC staff concluded in its TR SE that this change allows repairs of the inoperable low set relief valve to be performed in a plant operating mode with lower risks.
3.3.5 LCO 3.6.1.7: Reactor Building-to-Suppression Chamber Vacuum Breakers The reactor building-to-suppression chamber vacuum breakers relieve vacuum when the primary containment depressurizes below the pressure of the reactor building, thereby serving to preserve the integrity of the primary containment.
The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.
3.3.5   LCO 3.6.1.7: Reactor Building-to-Suppression Chamber Vacuum Breakers The reactor building-to-suppression chamber vacuum breakers relieve vacuum when the primary containment depressurizes below the pressure of the reactor building, thereby serving to preserve the integrity of the primary containment.
LCO: Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE.
LCO: Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE.
Condition Requiring Entry into End State: If one line has one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening, the breaker(s) must be returned to operability within 7 days (Required Action C.1). If the vacuum breaker(s) cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2). Modification for End State Required Actions: A new Condition D modifies the Required Actions so that if the vacuum breaker(s) cannot be returned to operable status within the required Completion Times, the plant is placed in hot shutdown.
Condition Requiring Entry into End State: If one line has one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening, the breaker(s) must be returned to operability within 7 days (Required Action C.1). If the vacuum breaker(s) cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2).
A !'Jote is added to the TS Required Action D.1 stating that LCO 3.0A.a is not applicable when entering Mode 3. Existing Conditions D and E are renumbered to E and F without changing the Required Action end state. Condition F would require shutting down the plant to Mode 3 (Required Action F.1) and then Mode 4 (Required Action F.2), to address an inability to comply with the required actions related to Conditions A, B, and E. Assessment:
Modification for End State Required Actions: A new Condition D modifies the Required Actions so that if the vacuum breaker(s) cannot be returned to operable status within the required Completion Times, the plant is placed in hot shutdown. A !'Jote is added to the TS Required Action D.1 stating that LCO 3.0A.a is not applicable when entering Mode 3. Existing Conditions D and E are renumbered to E and F without changing the Required Action end state. Condition F would require shutting down the plant to Mode 3 (Required Action F.1) and then Mode 4 (Required Action F.2), to address an inability to comply with the required actions related to Conditions A, B, and E.
In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The NRC staff evaluated this conclusion in its TR SE. The
Assessment: In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The NRC staff evaluated this conclusion in its TR SE. The
-reduced end state would only be applicable to the situation where the vacuum breaker(s) in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function.
 
The existing end state remains unchanged, as established by new Condition F, for conditions involving more than one inoperable line or vacuum breaker, since they are needed in Modes 1, 2, and 3. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3. In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling. The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
                                              - 13 reduced end state would only be applicable to the situation where the vacuum breaker(s) in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function. The existing end state remains unchanged, as established by new Condition F, for conditions involving more than one inoperable line or vacuum breaker, since they are needed in Modes 1, 2, and 3. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling.
3.3.6 LCD 3.6.1.8: Suppression Chamber-to-Drywell Vacuum Breakers The function of the suppression chamber-to-drywell vacuum breakers is to relieve vacuum in the drywell, thereby preventing an excessive negative differential pressure across the wetwell/drywell boundary.
Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3. In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling.
LCD: Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening, AND 12 suppression chamber-to-drywell vacuum breakers shall be closed. Condition Requiring Entry into End State: If one suppression chamber-to-drywell vacuum breaker is inoperable for opening, the vacuum breaker must be returned to operability within 72 hours (Required Action A.1). If the vacuum breaker cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action C.1) and in Mode 4 within 36 hours (Required Action C.2). Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if vacuum breaker(s) cannot be returned to operable status within the required Completion Time, the plant is placed in hot shutdown.
The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band C are renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) and Mode 4 (Required Action D.2), to address an inability to comply with the required actions related to Condition C. Assessment:
3.3.6   LCD 3.6.1.8: Suppression Chamber-to-Drywell Vacuum Breakers The function of the suppression chamber-to-drywell vacuum breakers is to relieve vacuum in the drywell, thereby preventing an excessive negative differential pressure across the wetwell/drywell boundary.
In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The reduced end state would only be applicable to the situation where one or more vacuum breakers in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function.
LCD: Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening, AND 12 suppression chamber-to-drywell vacuum breakers shall be closed.
In the NRC staff's TR SE, the staff concluded that by remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, maintaining defense-in-depth.
Condition Requiring Entry into End State: If one suppression chamber-to-drywell vacuum breaker is inoperable for opening, the vacuum breaker must be returned to operability within 72 hours (Required Action A.1). If the vacuum breaker cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action C.1) and in Mode 4 within 36 hours (Required Action C.2).
In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling.
Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if vacuum breaker(s) cannot be returned to operable status within the required Completion Time, the plant is placed in hot shutdown. A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band C are renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) and Mode 4 (Required Action D.2), to address an inability to comply with the required actions related to Condition C.
-The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Assessment: In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The reduced end state would only be applicable to the situation where one or more vacuum breakers in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function.
3.3.7 LCO 3.6.2.3: RHR Suppression Pool Cooling Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by two redundant RHR suppression pool cooling subsystems.
In the NRC staff's TR SE, the staff concluded that by remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, maintaining defense-in-depth. In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling.
 
                                                - 14 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.7   LCO 3.6.2.3: RHR Suppression Pool Cooling Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by two redundant RHR suppression pool cooling subsystems.
LCO: Two RHR suppression pool cooling subsystems shall be OPERABLE.
LCO: Two RHR suppression pool cooling subsystems shall be OPERABLE.
Condition Requiring Entry into End State: If one RHR suppression pool cooling subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If two RHR suppression pool cooling subsystems are inoperable (Condition B), one RHR suppression pool cooling system must be restored to operable status within 8 hours (Required Action B.1). If the RHR suppression pool cooling subsystem cannot be restored to operable status within the allotted time (Condition C), the plant must be placed in Mode 3 within 12 hours (Required Action C.1), and in Mode 4 within 36 hours (Required Action C.2). Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if the Required Action and associated Completion Time of Condition A are not met, the plant is placed in hot shutdown (Mode 3). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band Care renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) after 12 hours, and to Mode 4 (Required Action D.2) after 36 hours, to address an inability to comply with the required actions related to Condition C. Assessment:
Condition Requiring Entry into End State: If one RHR suppression pool cooling subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If two RHR suppression pool cooling subsystems are inoperable (Condition B), one RHR suppression pool cooling system must be restored to operable status within 8 hours (Required Action B.1). If the RHR suppression pool cooling subsystem cannot be restored to operable status within the allotted time (Condition C), the plant must be placed in Mode 3 within 12 hours (Required Action C.1), and in Mode 4 within 36 hours (Required Action C.2).
In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 end state was completed.
Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if the Required Action and associated Completion Time of Condition A are not met, the plant is placed in hot shutdown (Mode 3). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band Care renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) after 12 hours, and to Mode 4 (Required Action D.2) after 36 hours, to address an inability to comply with the required actions related to Condition C.
The results described in the BWROG TR, and evaluated by the NRC staff in its TR SE, indicated that the core damage risks while operating in Mode 3 (assuming the individual failure conditions) are lower or comparable to the current end state. For QCNPS, one loop of the suppression pool cooling system is sufficient to accomplish the required safety function.
Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 end state was completed. The results described in the BWROG TR, and evaluated by the NRC staff in its TR SE, indicated that the core damage risks while operating in Mode 3 (assuming the individual failure conditions) are lower or comparable to the current end state. For QCNPS, one loop of the suppression pool cooling system is sufficient to accomplish the required safety function. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for RCS makeup and cooling.
By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for RCS makeup and cooling. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.8 LCO 3.6.2.4: RHR Suppression Pool Spray Following a design-basis accident (DBA), the RHR suppression pool spray system removes heat from the suppression chamber airspace.
3.3.8   LCO 3.6.2.4: RHR Suppression Pool Spray Following a design-basis accident (DBA), the RHR suppression pool spray system removes heat from the suppression chamber airspace. A minimum of one RHR suppression pool spray subsystem is required to mitigate potential bypass leakage paths from the drywell and maintain the primary containment peak pressure below the design limits.
A minimum of one RHR suppression pool spray subsystem is required to mitigate potential bypass leakage paths from the drywell and maintain the primary containment peak pressure below the design limits.
 
-15LCO: Two RHR suppression pool spray subsystems shall be OPERABLE.
                                                  - 15 LCO: Two RHR suppression pool spray subsystems shall be OPERABLE.
Condition Requiring Entry into End State: If one RHR suppression pool spray subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If both RHR suppression pool spray subsystems are inoperable (Condition B), one of them must be restored to operable status within 8 hours (Required Action B.1). If the RHR suppression pool spray subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action C.1), and in Mode 4 within 36 hours (Required Action C.2). Modification for End State Required Actions: Required Action C.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action C.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
Condition Requiring Entry into End State: If one RHR suppression pool spray subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If both RHR suppression pool spray subsystems are inoperable (Condition B), one of them must be restored to operable status within 8 hours (Required Action B.1). If the RHR suppression pool spray subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action C.1), and in Mode 4 within 36 hours (Required Action C.2).
The main function of the RHR suppression spray system is to remove heat from the suppression chamber so that the pressure and temperature inside primary containment remain within analyzed design limits. The RHR suppression spray system was designed to mitigate potential effects of a postulated DBA, that is, a large LOCA which is assumed to occur concurrently with the most limiting single failure, and assuming conservative inputs. Under the conditions assumed in the DBA, steam blown down from the break could bypass the suppression pool and end up in the suppression chamber air space. The RHR suppression spray system could be needed to condense such steam so that the pressure and temperature inside primary containment remain within analyzed design limits. However, the frequency of a DBA is very small and the containment has considerable margin to failure above the design limits. For these reasons, the unavailability of one or both RHR suppression spray subsystems has no significant impact on CDF or LERF, even for accidents initiated during operation at power. Therefore, it is very unlikely that the RHR suppression spray system will be challenged to mitigate an accident occurring during power operation.
Modification for End State Required Actions: Required Action C.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action C.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
This probability becomes extremely unlikely for accidents that would occur during a small fraction of the year (less than 3 days) during which the plant would be in Mode 3 (associated with lower initial energy level and reduced decay heat load as compared to power operation) to repair the failed RHR suppression spray system. Section 6 of the NRC staff's TR SE summarizes the staff's risk argument for approval of the BWROG TR's Section 4.5.1.11 and LCO 3.6.2.4. The argument for staying in Mode 3 instead of going to Mode 4 to repair the RHR Suppression Pool Spray system (one or both trains) is also supported by defense-in-depth considerations.
Assessment: The main function of the RHR suppression spray system is to remove heat from the suppression chamber so that the pressure and temperature inside primary containment remain within analyzed design limits. The RHR suppression spray system was designed to mitigate potential effects of a postulated DBA, that is, a large LOCA which is assumed to occur concurrently with the most limiting single failure, and assuming conservative inputs. Under the conditions assumed in the DBA, steam blown down from the break could bypass the suppression pool and end up in the suppression chamber air space. The RHR suppression spray system could be needed to condense such steam so that the pressure and temperature inside primary containment remain within analyzed design limits.
TR SE Section 5.2 of the NRC staff's SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases, and precludes the need for RHR suppression pool spray subsystems.
However, the frequency of a DBA is very small and the containment has considerable margin to failure above the design limits. For these reasons, the unavailability of one or both RHR suppression spray subsystems has no significant impact on CDF or LERF, even for accidents initiated during operation at power. Therefore, it is very unlikely that the RHR suppression spray system will be challenged to mitigate an accident occurring during power operation. This probability becomes extremely unlikely for accidents that would occur during a small fraction of the year (less than 3 days) during which the plant would be in Mode 3 (associated with lower initial energy level and reduced decay heat load as compared to power operation) to repair the failed RHR suppression spray system.
In addition, the probability of a DBA (large-break) is much smaller during shutdown as compared to power operation.
Section 6 of the NRC staff's TR SE summarizes the staff's risk argument for approval of the BWROG TR's Section 4.5.1.11 and LCO 3.6.2.4. The argument for staying in Mode 3 instead of going to Mode 4 to repair the RHR Suppression Pool Spray system (one or both trains) is also supported by defense-in-depth considerations. TR SE Section 5.2 of the NRC staff's SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases, and precludes the need for RHR suppression pool spray subsystems. In addition, the probability of a DBA (large-break) is much smaller during shutdown as compared to power operation. A DBA in Mode 3 would be considerably less severe than a DBA occurring during power operation since Mode 3 is associated with lower initial energy level and reduced decay heat load. Under these extremely unlikely conditions, an alternate method that can be used to remove heat from the primary containment (in order to keep the pressure and temperature within the analyzed design basis limits) is containment venting. For more realistic accidents that could occur in Mode 3, several alternate means are available to remove heat from the primary containment, such as the RHR system in the suppression pool cooling
A DBA in Mode 3 would be considerably less severe than a DBA occurring during power operation since Mode 3 is associated with lower initial energy level and reduced decay heat load. Under these extremely unlikely conditions, an alternate method that can be used to remove heat from the primary containment (in order to keep the pressure and temperature within the analyzed design basis limits) is containment venting. For more realistic accidents that could occur in Mode 3, several alternate means are available to remove heat from the primary containment, such as the RHR system in the suppression pool cooling
 
-mode and the containment spray mode. The risk and defense-in-depth arguments used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable RHR suppression pool spray system. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
                                                - 16 mode and the containment spray mode. The risk and defense-in-depth arguments used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable RHR suppression pool spray system.
3.3.9 LCO 3.6.4.1: Secondary Containment Following a DBA, the function of the secondary containment is to contain, dilute, and stop radioactivity (mostly fission products) that may leak from primary containment.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Its leak tightness is required to ensure that the release of radioactivity from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the standby gas treatment (SGT) system prior to discharge to the environment.
3.3.9   LCO 3.6.4.1: Secondary Containment Following a DBA, the function of the secondary containment is to contain, dilute, and stop radioactivity (mostly fission products) that may leak from primary containment. Its leak tightness is required to ensure that the release of radioactivity from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the standby gas treatment (SGT) system prior to discharge to the environment.
LCO: The secondary containment shall be OPERABLE.
LCO: The secondary containment shall be OPERABLE.
Condition Requiring Entry into End State: If the secondary containment is inoperable, it must be restored to operable status within 4 hours (Required Action A.1). If it cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action B.1), and in Mode 4 within 36 hours (Required Action B.2). Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
Condition Requiring Entry into End State: If the secondary containment is inoperable, it must be restored to operable status within 4 hours (Required Action A.1). If it cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action B.1), and in Mode 4 within 36 hours (Required Action B.2).
This LCO entry condition does not include gross leakage through an unisolable release path. In the BWROG TR, it was concluded that previous generic PRA work related to 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," requirements has shown that containment leakage is not risk significant.
Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
The primary containment and all other primary and secondary containment-related functions would still be operable, including the SGT system, thereby minimizing the likelihood of an unacceptable release. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared with transitioning to Mode 4. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Assessment: This LCO entry condition does not include gross leakage through an unisolable release path. In the BWROG TR, it was concluded that previous generic PRA work related to 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," requirements has shown that containment leakage is not risk significant. The primary containment and all other primary and secondary containment-related functions would still be operable, including the SGT system, thereby minimizing the likelihood of an unacceptable release. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared with transitioning to Mode 4.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.10 LCO 3.6.4.3: Standby Gas Treatment System The function of the SGT system is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a DBA are filtered and adsorbed prior to exhausting to the environment.
3.3.10 LCO 3.6.4.3: Standby Gas Treatment System The function of the SGT system is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a DBA are filtered and adsorbed prior to exhausting to the environment.
LCO: Two SGT subsystems shall be OPERABLE.
LCO: Two SGT subsystems shall be OPERABLE.
-17Condition Requiring Entrv into End State: If one SGT subsystem is inoperable, it must be restored to operable status within 7 days (Required Action A.1). If the SGT subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action B.1) and in Mode 4 within 36 hours (Required Action B.2). In addition, if two SGT subsystems are inoperable in Mode 1, 2, or 3, (Condition D), and one SGT system can not be restored to operable within 1 hour then the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2). Modification for End State Required Actions: Required Action B.2 and E.2 are deleted allowing the plant to stay in l\IIode 3 while completing repairs. A Note is added to the TS Required Actions B.1 and E.1 stating that LCO 3.0A.a is not applicable when entering Mode 3. Assessment:
 
The unavailability of one or both SGT subsystems has no impact on CDF or LERF, irrespective of the mode of operation at the time of the accident.
                                                - 17 Condition Requiring Entrv into End State: If one SGT subsystem is inoperable, it must be restored to operable status within 7 days (Required Action A.1). If the SGT subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours (Required Action B.1) and in Mode 4 within 36 hours (Required Action B.2). In addition, if two SGT subsystems are inoperable in Mode 1, 2, or 3, (Condition D), and one SGT system can not be restored to operable within 1 hour then the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2).
Furthermore, the challenge frequency of the SGT system (i.e., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases resulting from materials that leak from the primary to the secondary containment above TS limits) is less than 1.0E-6/yr.
Modification for End State Required Actions: Required Action B.2 and E.2 are deleted allowing the plant to stay in l\IIode 3 while completing repairs. A Note is added to the TS Required Actions B.1 and E.1 stating that LCO 3.0A.a is not applicable when entering Mode 3.
Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (i.e., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release. Section 5.2 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases.
Assessment: The unavailability of one or both SGT subsystems has no impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the SGT system (i.e., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases resulting from materials that leak from the primary to the secondary containment above TS limits) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (i.e., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.
Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Sections 4.5.1.13 and 4.5.2.11, and LCO 3.6.4.3. According to this evaluation, which applies to BWR plant types, (such as, QCNPS's BWR-3 design), staying in Mode 3 instead of going to Mode 4 to repair the SGT system (one or both trains) is also supported by defense-in-depth considerations.
Section 5.2 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Sections 4.5.1.13 and 4.5.2.11, and LCO 3.6.4.3. According to this evaluation, which applies to BWR plant types, (such as, QCNPS's BWR-3 design), staying in Mode 3 instead of going to Mode 4 to repair the SGT system (one or both trains) is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable SGT system.
The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable SGT system. The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Actions B.1 and E.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.
The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Actions B.1 and E.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.
3.3.11 LCO 3.7.1: Residual Heat Removal Service Water (RHRSW) System The RHRSW system is designed to provide cooling water for the RHR system heat exchangers, which are required for safe shutdown following a normal shutdown or DBA or transient.
3.3.11 LCO 3.7.1: Residual Heat Removal Service Water (RHRSW) System The RHRSW system is designed to provide cooling water for the RHR system heat exchangers, which are required for safe shutdown following a normal shutdown or DBA or transient.
LCO: Two RHRSW subsystems shall be OPERABLE.
LCO: Two RHRSW subsystems shall be OPERABLE.
Condition Requiring Entrv into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions: If one RHRSW pump is inoperable (Condition A), it must be restored to operable status within 30 days (Required Action A.1).
Condition Requiring Entrv into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions:
-b. If one RHRSW pump in each subsystem is inoperable (Condition B), one RHRSW pump must be restored to operable status within 7 days (Required Action B.1). c. If one RHRSW subsystem is inoperable for reasons other than Condition A (Condition C), the RHRSW subsystem must be restored to operable status within 7 days (Required Action C.1). d. If both RHRSW subsystems are inoperable for reasons other than Condition B (Condition D), then one RHRSW subsystem must be restored to operable within 8 hours. If the required action and associated completion time cannot be met within the allotted time (Condition E), the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2) Modification for End State Required Actions: A new Condition D is added which establishes requirements for existing Conditions A, B, and C, that are similar to existing Condition E, but without Required Action E.2. A Note is added to the Required Actions D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. The existing Required Conditions D and E are renumbered to Conditions E and F, respectively.
: a.        If one RHRSW pump is inoperable (Condition A), it must be restored to operable status within 30 days (Required Action A.1).
Required Condition F now applies to Condition E, which maintains the existing requirements with respect to both RHRSW subsystems being inoperable for reasons other than Condition B. Assessment:
 
In the BWROG TR, a comparative PRA evaluation of the core damage risks when operating in the current end state versus the Mode 3 end state was performed.
                                                - 18
The NRC staff conclusions, discussed in its TR SE, indicate that the core damage risks are lower when remaining in Mode 3 compared with transitioning to the current end state of Mode 4. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray are needed for RCS makeup and cooling. Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3, and the required safety function can still be performed with the RHRSW subsystem components that are still operable.
: b.       If one RHRSW pump in each subsystem is inoperable (Condition B), one RHRSW pump must be restored to operable status within 7 days (Required Action B.1).
: c.       If one RHRSW subsystem is inoperable for reasons other than Condition A (Condition C), the RHRSW subsystem must be restored to operable status within 7 days (Required Action C.1).
: d.       If both RHRSW subsystems are inoperable for reasons other than Condition B (Condition D), then one RHRSW subsystem must be restored to operable within 8 hours.
: e.        If the required action and associated completion time cannot be met within the allotted time (Condition E), the plant must be placed in Mode 3 within 12 hours (Required Action E.1) and in Mode 4 within 36 hours (Required Action E.2)
Modification for End State Required Actions: A new Condition D is added which establishes requirements for existing Conditions A, B, and C, that are similar to existing Condition E, but without Required Action E.2. A Note is added to the Required Actions D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. The existing Required Conditions D and E are renumbered to Conditions E and F, respectively. Required Condition F now applies to Condition E, which maintains the existing requirements with respect to both RHRSW subsystems being inoperable for reasons other than Condition B.
Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks when operating in the current end state versus the Mode 3 end state was performed. The NRC staff conclusions, discussed in its TR SE, indicate that the core damage risks are lower when remaining in Mode 3 compared with transitioning to the current end state of Mode 4. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray are needed for RCS makeup and cooling. Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3, and the required safety function can still be performed with the RHRSW subsystem components that are still operable.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions D.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions D.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.12 LCO 3.7.4: Control Room Emergency Ventilation (CREV) system The CREV system provides a radiologically controlled environment from which the plant can be safely operated following a DBA. LCO: The CREV system shall be OPERABLE.
3.3.12 LCO 3.7.4: Control Room Emergency Ventilation (CREV) system The CREV system provides a radiologically controlled environment from which the plant can be safely operated following a DBA.
Condition Requiring Entry into End State: If the CREV system is inoperable in Mode 1,2, or 3 for reasons other than an inoperable control room envelope (CRE) boundary in Mode 1, 2, or 3, (Condition B), it must be restored to operable status within 7 days (Required Action A.1). If the
LCO: The CREV system shall be OPERABLE.
-CREV system is inoperable in Mode 1, 2, or 3 due to inoperable Control Room Envelope boundary, the licensee must initiate mitigating actions immediately (Required Action B.1), verify mitigating actions ensure CRE occupant exposures to airborne hazards will not exceed limits within 24 hours (Required Actions B.2), and restore CRE operability within 90 days (Required Actions B.3). If the CREV system cannot be restored to operable status within the allotted time for Condition A or B, the plant must be placed in Mode 3 within 12 hours (Required Action C.1) and in Mode 4 within 36 hours (Required Action Co2). Modification for End State Required Actions: The change adds a new Condition B with Required Action B.1 to be in Mode 3 within 12 hours when Required Action and associated Completion Time of Condition A are not met in Mode 1, 2, or 3. The change renumbers old Conditions B, C and D to Conditions C, D and E. The renumbered Condition D would require that if the CREV system cannot be restored to operable status within the allotted time for renumbered Condition C, the plant must be placed in Mode 3 within 12 hours (Required Action D.1) and in Mode 4 within 36 hours (Required Action D.2). A Note is added to the new Required Actions B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
Condition Requiring Entry into End State: If the CREV system is inoperable in Mode 1,2, or 3 for reasons other than an inoperable control room envelope (CRE) boundary in Mode 1, 2, or 3, (Condition B), it must be restored to operable status within 7 days (Required Action A.1). If the
The unavailability of the CREV system has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident.
 
Furthermore, the challenge frequency of the CREV system (Le., the frequency with which the system is expected to be challenged to provide a radiologically-controlled environment in the main control room following a DBA which leads to core damage and leaks of radiation from the containment that can reach the control room) is less than 1.0E-6/yr.
                                                - 19 CREV system is inoperable in Mode 1, 2, or 3 due to inoperable Control Room Envelope boundary, the licensee must initiate mitigating actions immediately (Required Action B.1), verify mitigating actions ensure CRE occupant exposures to airborne hazards will not exceed limits within 24 hours (Required Actions B.2), and restore CRE operability within 90 days (Required Actions B.3). If the CREV system cannot be restored to operable status within the allotted time for Condition A or B, the plant must be placed in Mode 3 within 12 hours (Required Action C.1) and in Mode 4 within 36 hours (Required Action Co2).
Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release. Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases.
Modification for End State Required Actions: The change adds a new Condition B with Required Action B.1 to be in Mode 3 within 12 hours when Required Action and associated Completion Time of Condition A are not met in Mode 1, 2, or 3. The change renumbers old Conditions B, C and D to Conditions C, D and E. The renumbered Condition D would require that if the CREV system cannot be restored to operable status within the allotted time for renumbered Condition C, the plant must be placed in Mode 3 within 12 hours (Required Action D.1) and in Mode 4 within 36 hours (Required Action D.2). A Note is added to the new Required Actions B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.16, and LCO 3.7.4, "Main Control Room Environmental Control (MCREC) System." (Note: The CREV system at QCNPS serves a similar design purpose as the MCREC described in NUREG 1433). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by defense-in-depth considerations.
Assessment: The unavailability of the CREV system has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the CREV system (Le., the frequency with which the system is expected to be challenged to provide a radiologically-controlled environment in the main control room following a DBA which leads to core damage and leaks of radiation from the containment that can reach the control room) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.
The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable CREV system. The Note "LCO LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.16, and LCO 3.7.4, "Main Control Room Environmental Control (MCREC) System." (Note: The CREV system at QCNPS serves a similar design purpose as the MCREC described in NUREG 1433). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable CREV system.
The Note "LCO LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.13 LCO 3.7.5: Control Room Emergency Ventilation Air Conditioning System The Control Room AC system provides temperature control for the control room following control room isolation during accident conditions.
3.3.13 LCO 3.7.5: Control Room Emergency Ventilation Air Conditioning System The Control Room AC system provides temperature control for the control room following control room isolation during accident conditions.
LCO: The Control Room Emergency Ventilation Air Conditioning System shall be OPERABLE.
LCO: The Control Room Emergency Ventilation Air Conditioning System shall be OPERABLE.
-Condition Requiring Entry into End State: If the Control Room Emergency Ventilation Air Conditioning System is inoperable, the system must be restored to operable status within 30 days (Required Action A.1). If the required actions and associated completion times for Condition A cannot be met, the plant must be placed in Mode 3 within 12 hours (Required Action B.1) and in Mode 4 within 36 hours (Required Action B.2). Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
 
The unavailability of the Control Room Emergency Ventilation Air Conditioning System has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident.
                                                - 20 Condition Requiring Entry into End State: If the Control Room Emergency Ventilation Air Conditioning System is inoperable, the system must be restored to operable status within 30 days (Required Action A.1). If the required actions and associated completion times for Condition A cannot be met, the plant must be placed in Mode 3 within 12 hours (Required Action B.1) and in Mode 4 within 36 hours (Required Action B.2).
Furthermore, the challenge frequency of the system (Le., the frequency with which the system is expected to be challenged to provide temperature control for the control room following control room isolation following a DBA) is less than 1.0E-6/yr.
Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively), is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release. Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases.
Assessment: The unavailability of the Control Room Emergency Ventilation Air Conditioning System has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the system (Le., the frequency with which the system is expected to be challenged to provide temperature control for the control room following control room isolation following a DBA) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively), is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.
Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.17, and LCO 3.7.5, "Control Room Air Conditioning (AC) System," (which functions similar to QCNPS's Control Room Emergency Ventilation Air Conditioning System). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by depth considerations.
Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.17, and LCO 3.7.5, "Control Room Air Conditioning (AC) System," (which functions similar to QCNPS's Control Room Emergency Ventilation Air Conditioning System). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by defense-in depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable Control Room Emergency Ventilation Air Conditioning System.
The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable Control Room Emergency Ventilation Air Conditioning System. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.14 LCO 3.7.6: Main Condenser Offgas The offgas from the main condenser normally includes radioactive gases. The gross gamma activity rate is controlled to ensure that accident analysis assumptions are satisfied and that offsite dose limits will not be exceeded during postulated accidents.
3.3.14 LCO 3.7.6: Main Condenser Offgas The offgas from the main condenser normally includes radioactive gases. The gross gamma activity rate is controlled to ensure that accident analysis assumptions are satisfied and that offsite dose limits will not be exceeded during postulated accidents. The Main Condenser Offgas (MCOG) gross gamma activity rate is an initial condition of a DBA which assumes a gross failure of the MCOG system pressure boundary.
The Main Condenser Offgas (MCOG) gross gamma activity rate is an initial condition of a DBA which assumes a gross failure of the MCOG system pressure boundary.
LCO: The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be S 251,100 IJCi/second after decay of 30 minutes.
LCO: The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be S 251,100 IJCi/second after decay of 30 minutes.
 
-21 Condition Requirinq Entry into End State: If the gross gamma activity rate of the noble gases in the MCOG system is not within limits, the gross gamma activity rate of the noble gases in the main condenser Offgas must be restored to within limits within 72 hours (Required Action A.1). If the required action and associated completion time cannot be met, one of the following must occur: a. All main steamlines must be isolated within 12 hours (Required Action B.1). b. The steam jet air ejector must be isolated within 12 hours (Required Action B.2). c. The plant must be placed in Mode 3 within 12 hours (Required Action B.3.1) and in Mode 4 within 36 hours (Required Action 8.3.2). Modification for End State Required Actions: Required Action B.3.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. Required Action B.3.1 is renumbered to Required Action B.3 and a Note is added to the TS Required Action B.3 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
                                                - 21 Condition Requirinq Entry into End State: If the gross gamma activity rate of the noble gases in the MCOG system is not within limits, the gross gamma activity rate of the noble gases in the main condenser Offgas must be restored to within limits within 72 hours (Required Action A.1).
The failure to maintain the gross gamma activity rate of the noble gases in the MCOG within limits has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident.
If the required action and associated completion time cannot be met, one of the following must occur:
Furthermore, the challenge frequency of the MCOG system (Le., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases following a DBA) is less than 1.0E-6/yr.
: a.       All main steamlines must be isolated within 12 hours (Required Action B.1).
Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release. Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases.
: b.       The steam jet air ejector must be isolated within 12 hours (Required Action B.2).
Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.18 and LCO 3.7.6. The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the MCOG system (one or both trains) is also supported by defense-in-depth considerations.
: c.       The plant must be placed in Mode 3 within 12 hours (Required Action B.3.1) and in Mode 4 within 36 hours (Required Action 8.3.2).
The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable MCOG system. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.3.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Modification for End State Required Actions: Required Action B.3.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. Required Action B.3.1 is renumbered to Required Action B.3 and a Note is added to the TS Required Action B.3 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
3.3.15 LCO 3.8.1: Alternating Current (AC) Sources -Operating The purpose of the AC electrical system is to provide the power required to put and maintain the plant in a safe condition and prevent the release of radioactivity to the environment.
Assessment: The failure to maintain the gross gamma activity rate of the noble gases in the MCOG within limits has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the MCOG system (Le., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases following a DBA) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.
The Class 1E electrical power distribution system AC sources consist of the offsite power source (preferred power sources, normal and alternate(s)), and the onsite standby power sources (e.g., diesel generators (DGs)).
Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.18 and LCO 3.7.6. The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the MCOG system (one or both trains) is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable MCOG system.
-Per the QCNPS Updated Final Safety Analysis Report, the design of the AC electrical system provides independence and redundancy.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.3.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
The onsite Class 1E AC distribution system is divided into redundant divisions so that the loss of anyone division does not prevent the minimum safety functions from being performed.
3.3.15 LCO 3.8.1: Alternating Current (AC) Sources - Operating The purpose of the AC electrical system is to provide the power required to put and maintain the plant in a safe condition and prevent the release of radioactivity to the environment. The Class 1E electrical power distribution system AC sources consist of the offsite power source (preferred power sources, normal and alternate(s)), and the onsite standby power sources (e.g., diesel generators (DGs)).
Each division has connections to two preferred offsite power sources and a single DG or other Class 1E Standby AC power source. Offsite power at QCNPS is supplied to the unit switchyard(s) from the transmission network by five 345-kV transmission lines. Electric energy generated at the station is stepped up to 345-kV by the main power transformers and fed into the station's 345-kV transmission terminal ring bus which is also connected to the five transmission lines. The 345-kV ring bus provides power to two reserve auxiliary power transformers.
 
Each reserve auxiliary transformer has sufficient capacity to handle the auxiliary power requirements of one unit. Each of these auxiliary power supplies is available, through circuit breaker switching, to both Division I and the Division II emergency auxiliary equipment of both units, and therefore, serves as a redundant offsite source of auxiliary power. In the event of a loss of offsite power, the emergency electrical loads are automatically connected to the operating DGs in sufficient time to provide for a safe reactor shutdown and to mitigate the consequences of a DBA. LCO: The following AC electrical power sources shall be OPERABLE:
                                                - 22 Per the QCNPS Updated Final Safety Analysis Report, the design of the AC electrical system provides independence and redundancy. The onsite Class 1E AC distribution system is divided into redundant divisions so that the loss of anyone division does not prevent the minimum safety functions from being performed. Each division has connections to two preferred offsite power sources and a single DG or other Class 1E Standby AC power source.
: a. Two qualified circuits between the offsite transmission network and the onsite Class1 E AC Electric Power Distribution System. b. Two DGs One qualified circuit between the offsite transmission network and the opposite unit's onsite Class 1E AC Electrical Power Distribution System capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5. (Unit 2 only); and The opposite unit's DG capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5 (Unit 2 only). Condition Requiring Entry into End State: Plant operators must bring the plant to Mode 3 within 12 hours and Mode 4 within 36 hours following the sustained inoperability of either or both required offsite circuits; one or two required DGs; or one required offsite circuit and one required DG. Modification for End State Required Actions: Required Action F.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. The plant will remain in Mode 3 (hot shutdown) (Required Action F.1). A Note is added to the TS Required Action F.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
Offsite power at QCNPS is supplied to the unit switchyard(s) from the transmission network by five 345-kV transmission lines. Electric energy generated at the station is stepped up to 345-kV by the main power transformers and fed into the station's 345-kV transmission terminal ring bus which is also connected to the five transmission lines. The 345-kV ring bus provides power to two reserve auxiliary power transformers. Each reserve auxiliary transformer has sufficient capacity to handle the auxiliary power requirements of one unit. Each of these auxiliary power supplies is available, through circuit breaker switching, to both Division I and the Division II emergency auxiliary equipment of both units, and therefore, serves as a redundant offsite source of auxiliary power. In the event of a loss of offsite power, the emergency electrical loads are automatically connected to the operating DGs in sufficient time to provide for a safe reactor shutdown and to mitigate the consequences of a DBA.
Entry into any of the conditions for the AC power sources implies that the AC power sources have been degraded and the single failure protection for the safe shutdown equipment may be ineffective.
LCO: The following AC electrical power sources shall be OPERABLE:
Consequently, as specified in the current STS 3.8.1, the plant operators must bring the plant to Mode 4 when the required action is not completed by the specified time for the associated action.
: a. Two qualified circuits between the offsite transmission network and the onsite Class1 E AC Electric Power Distribution System.
-In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed.
: b. Two DGs
Events initiated by the loss of offsite power are dominant contributors to core damage frequency in most BWR PRAs, and RCIC and HPCI play major roles in mitigating these events. In the !\JRC staft's TR SE, the NRC staff concluded that, based on the low probability of loss of the AC power and the number of systems available in Mode 3, the core damage risks are lower in Mode 3 than in Mode 4 for one inoperable AC power source. For QCNPS, going to Mode 4 for one inoperable AC power source would cause loss of RCIC and HPCI, and loss of the power conversion systems, and would require activating the RHR system. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action for F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
: c. One qualified circuit between the offsite transmission network and the opposite unit's onsite Class 1E AC Electrical Power Distribution System capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5. (Unit 2 only); and
3.3.16 LCO 3.8.4: Direct Current (DC) Sources -Operating The purpose of the DC power system is to provide a reliable source of DC power for both normal and abnormal conditions.
: d. The opposite unit's DG capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5 (Unit 2 only).
It must supply power in an emergency for an adequate length of time until normal supplies can be restored.
Condition Requiring Entry into End State: Plant operators must bring the plant to Mode 3 within 12 hours and Mode 4 within 36 hours following the sustained inoperability of either or both required offsite circuits; one or two required DGs; or one required offsite circuit and one required DG.
Modification for End State Required Actions: Required Action F.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. The plant will remain in Mode 3 (hot shutdown)
(Required Action F.1). A Note is added to the TS Required Action F.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
Assessment: Entry into any of the conditions for the AC power sources implies that the AC power sources have been degraded and the single failure protection for the safe shutdown equipment may be ineffective. Consequently, as specified in the current STS 3.8.1, the plant operators must bring the plant to Mode 4 when the required action is not completed by the specified time for the associated action.
 
                                                - 23 In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. Events initiated by the loss of offsite power are dominant contributors to core damage frequency in most BWR PRAs, and RCIC and HPCI play major roles in mitigating these events. In the !\JRC staft's TR SE, the NRC staff concluded that, based on the low probability of loss of the AC power and the number of systems available in Mode 3, the core damage risks are lower in Mode 3 than in Mode 4 for one inoperable AC power source. For QCNPS, going to Mode 4 for one inoperable AC power source would cause loss of RCIC and HPCI, and loss of the power conversion systems, and would require activating the RHR system.
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action for F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.16 LCO 3.8.4: Direct Current (DC) Sources - Operating The purpose of the DC power system is to provide a reliable source of DC power for both normal and abnormal conditions. It must supply power in an emergency for an adequate length of time until normal supplies can be restored.
LCO: For Modes 1, 2, and 3, the following DC electrical power subsystems shall be OPERABLE:
LCO: For Modes 1, 2, and 3, the following DC electrical power subsystems shall be OPERABLE:
: a. Two 250 VDC electrical power subsystems;
: a. Two 250 VDC electrical power subsystems;
: b. Division 1 and Division 2 125 VDC electrical power subsystems:
: b. Division 1 and Division 2 125 VDC electrical power subsystems: and
and c. The opposite unit's 125 VDC electrical power subsystem capable of supporting equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only), and LCO 3.8.1. Condition Requiring Entry into End State: The plant operators must bring the plant to Mode 3 within 12 hours (Required Action F.1) and Mode 4 within 36 hours (Required Action F.2) following the sustained inoperability of one 250 VDC electrical subsystem, or Division 1 or 2, DC electrical power subsystem for a period of 72 hours (Conditions A thru D), or 7 days if opposite unit 125 VDC electrical power subsystem is inoperable (Condition E). Modification for End State Required Actions: Required Action F.2 is deleted, allowing the plant to stay in Mode 3 while Conditions A thru E are not met. A !\Jote is added to the TS Required Action F.1 stating that LCO 3.0.4(a) is not applicable when entering Mode 3. Assessment:
: c. The opposite unit's 125 VDC electrical power subsystem capable of supporting equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only),
If one of the DC electrical power subsystems is inoperable, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition.
and LCO 3.8.1.
In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 and Mode 4 end states was performed, with one DC system inoperable.
Condition Requiring Entry into End State: The plant operators must bring the plant to Mode 3 within 12 hours (Required Action F.1) and Mode 4 within 36 hours (Required Action F.2) following the sustained inoperability of one 250 VDC electrical subsystem, or Division 1 or 2, DC electrical power subsystem for a period of 72 hours (Conditions A thru D), or 7 days if opposite unit 125 VDC electrical power subsystem is inoperable (Condition E).
Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the RCIC and HPCI systems play major roles in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause loss of the RCIC, HPCI and the power conversion systems (condenser/feedwater), and would require activating the RHR system.
Modification for End State Required Actions: Required Action F.2 is deleted, allowing the plant to stay in Mode 3 while Conditions A thru E are not met. A !\Jote is added to the TS Required Action F.1 stating that LCO 3.0.4(a) is not applicable when entering Mode 3.
-The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
Assessment: If one of the DC electrical power subsystems is inoperable, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 and Mode 4 end states was performed, with one DC system inoperable. Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the RCIC and HPCI systems play major roles in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4.
3.3.17 LCO 3.8.7: Distribution Systems -Operating The onsite Class 1E AC electrical power distribution system is divided into redundant and independent AC electrical power distribution systems. The primary AC electrical power distribution subsystem for each plant consists of two 4.16-kV Essential Service System (ESS) buses having an offsite source of power as well as a dedicated onsite DG source. The secondary plant distribution subsystems include 480-VAC ESS buses and associated load centers, motor control centers, distribution panels and transformers.
For QCNPS, going to Mode 4 would cause loss of the RCIC, HPCI and the power conversion systems (condenser/feedwater), and would require activating the RHR system.
The 120-VAC vital buses are arranged in three different subsystems:
 
120V reactor protection system, 120V instrumentation bus system, and 120V ESS. The 120V ESS bus is supplied by a static uninterruptible power supply. There are two independent 250 VDC station service electrical power distribution subsystems and two independent 125 VDC electrical power distribution subsystems.
                                                - 24 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
3.3.17 LCO 3.8.7: Distribution Systems - Operating The onsite Class 1E AC electrical power distribution system is divided into redundant and independent AC electrical power distribution systems. The primary AC electrical power distribution subsystem for each plant consists of two 4.16-kV Essential Service System (ESS) buses having an offsite source of power as well as a dedicated onsite DG source. The secondary plant distribution subsystems include 480-VAC ESS buses and associated load centers, motor control centers, distribution panels and transformers.
The 120-VAC vital buses are arranged in three different subsystems: 120V reactor protection system, 120V instrumentation bus system, and 120V ESS. The 120V ESS bus is supplied by a static uninterruptible power supply. There are two independent 250 VDC station service electrical power distribution subsystems and two independent 125 VDC electrical power distribution subsystems.
LCO: For Modes 1, 2, and 3, the following electrical power distribution subsystems shall be OPERABLE:
LCO: For Modes 1, 2, and 3, the following electrical power distribution subsystems shall be OPERABLE:
: a. Division 1 and Division 2 AC and DC electrical power distribution subsystems; and b. The portions of the opposite unit's AC and DC electrical power distribution subsystems necessary to support equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only), and LCO 3.8.1. Condition Requiring Entry into End State: The plant operators must bring the plant to Mode 3 within 12 hours and Mode 4 within 36 hours following the sustained inoperability of one or more AC, DC, or one or more required opposite unit AC or DC electrical power distribution subsystems inoperable for a period of 8 hours, 2 hours and 7 days, respectively (with a maximum 16 hour Completion Time limit from initial discovery of failure to meet the LCO, to preclude being in the LCO indefinitely).
: a. Division 1 and Division 2 AC and DC electrical power distribution subsystems; and
Modification for End State Required Actions: The TS change is to remove the requirement to place the plant in Mode 4 (Required Action D.2 is deleted).
: b. The portions of the opposite unit's AC and DC electrical power distribution subsystems necessary to support equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only), and LCO 3.8.1.
A Note is added to the TS Required Action D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Assessment:
Condition Requiring Entry into End State: The plant operators must bring the plant to Mode 3 within 12 hours and Mode 4 within 36 hours following the sustained inoperability of one or more AC, DC, or one or more required opposite unit AC or DC electrical power distribution subsystems inoperable for a period of 8 hours, 2 hours and 7 days, respectively (with a maximum 16 hour Completion Time limit from initial discovery of failure to meet the LCO, to preclude being in the LCO indefinitely).
If one of the AC/DC/AC ESS is inoperable, the remaining AC/DC/AC ESS have the capacity to support a safe shutdown and to mitigate an accident condition.
Modification for End State Required Actions: The TS change is to remove the requirement to place the plant in Mode 4 (Required Action D.2 is deleted). A Note is added to the TS Required Action D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed with one of the AC/DC/AC ESS inoperable.
Assessment: If one of the AC/DC/AC ESS is inoperable, the remaining AC/DC/AC ESS have the capacity to support a safe shutdown and to mitigate an accident condition. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed with one of the AC/DC/AC ESS inoperable. Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the high pressure core cooling systems, HPCI and RCIC, playa major role in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause a loss of the RCIC and HPCI systems, and the power conversion systems, and would require activating the RHR system.
Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the high pressure core cooling systems, HPCI and RCIC, playa major role in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause a loss of the RCIC and HPCI systems, and the power conversion systems, and would require activating the RHR system. The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1
The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1
-prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment. Overall Assessment of Proposed Technical changes: Based upon the above assessments, and because the time spent in Mode 3 to perform the repair on any of the systems described above would be infrequent and limited, and in light of defense-in-depth considerations (discussed above and in the BWROG TR, and as evaluated by the NRC staff's TR SE), the NRC staff concludes the changes to the QCNPS TSs described above are acceptable. STATE CONSULTATION In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment.
 
The State official had no comments. ENVIRONMENTAL CONSIDERATION The amendment changes requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
                                                    - 25 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.
The Commission has previously issued a proposed finding that adopting TSTF-423, Rev 0, involves no significant hazards considerations, and there has been no public comment on the finding in Federal Register Notice 70 FR 74037, December 14, 2005. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.  
3.4. Overall Assessment of Proposed Technical changes:
Based upon the above assessments, and because the time spent in Mode 3 to perform the repair on any of the systems described above would be infrequent and limited, and in light of defense-in-depth considerations (discussed above and in the BWROG TR, and as evaluated by the NRC staff's TR SE), the NRC staff concludes the changes to the QCNPS TSs described above are acceptable.
 
==4.0    STATE CONSULTATION==
 
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.
 
==5.0    ENVIRONMENTAL CONSIDERATION==
 
The amendment changes requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that adopting TSTF-423, Rev 0, involves no significant hazards considerations, and there has been no public comment on the finding in Federal Register Notice 70 FR 74037, December 14, 2005. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
 
==6.0    CONCLUSION==
 
The Commission has concluded, on the basis of the considerations discussed above, that (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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
 
==7.0    REFERENCES==
: 1.      NEDC-32988-A, Revision 2, "Technical Justification to Support Risk-Informed Modification to Selected Required Action End States for BWR Plants," December 2002.
(ADAMS Accession No. ML030170084).
: 2.      Federal Register, Vol. 58, No. 139, p. 39136, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Plants," July 22,1993 (58 FR 39132).
: 3.      10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
 
                                            - 26
: 4. Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000 (ADAMS Accession No. ML003699426).
: 5. Regulatory Guide 1.177, "An Approach for Plant Specific, Risk-Informed Decision Making: Technical Specifications," USNRC, August 1998. (ADAMS Accession 1\10.
ML003740176).
: 6. NRC Safety Evaluation for Topical Report NEDC-32988, Revision 2, September 27, 2002. (ADAMS Accession No. ML022700603).
: 7. NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 2, April 1996.
: 8. TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A." (ADAMS Accession No. ML032270250).
: 9. TSTF-IG-05-02, "Implementation Guidance for TSTF-423, Revision 0, 'Technical Specifications End States,' NEDC-32988-A," September 2005. (ADAMS Accession No. ML052700156).
: 10. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis,"
USNRC, August 1998. (ADAMS Accession No. ML003740133).
Principal Contributors: R. P. Grover, NRR Kristy Bucholtz, NRR Date: October 21, 2009


==6.0 CONCLUSION==
Mr. Charles G. Pardee                                        October 21,2009 President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555


The Commission has concluded, on the basis of the considerations discussed above, that (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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. REFERENCES NEDC-32988-A, Revision 2, "Technical Justification to Support Risk-Informed Modification to Selected Required Action End States for BWR Plants," December 2002. (ADAMS Accession No. ML030170084). Federal Register, Vol. 58, No. 139, p. 39136, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Plants," July 22,1993 (58 FR 39132). 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." 
==SUBJECT:==
-Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000 (ADAMS Accession No. ML003699426). Regulatory Guide 1.177, "An Approach for Plant Specific, Risk-Informed Decision Making: Technical Specifications," USNRC, August 1998. (ADAMS Accession 1\10. ML003740176). NRC Safety Evaluation for Topical Report NEDC-32988, Revision 2, September 27, 2002. (ADAMS Accession No. ML022700603). NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 2, April 1996. TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A." (ADAMS Accession No. ML032270250). TSTF-IG-05-02, "Implementation Guidance for TSTF-423, Revision 0, 'Technical Specifications End States,' NEDC-32988-A," September 2005. (ADAMS Accession No. ML052700156). Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis," USNRC, August 1998. (ADAMS Accession No. ML003740133).
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RISK-INFORMED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AND MD6998)
Principal R. P. Grover, NRR Kristy Bucholtz, NRR Date: October 21, 2009 Mr. Charles G. October 21,2009 President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: RISK-INFORMED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AND MD6998)  


==Dear Mr. Pardee:==
==Dear Mr. Pardee:==
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively.
 
The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession No. ML090350151).
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS)
The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2." A copy of the related Safety Evaluation is also enclosed.
Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession No. ML090350151).
The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRA! Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265  
The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2."
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, IRA!
Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265


==Enclosures:==
==Enclosures:==
: 1. Amendment No. to DPR-29 2. Amendment No. to DPR-30 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
: 1. Amendment No.           to DPR-29
PUBLIC RidsOgcRp Resource LPL3-2 RlF RidsNrrLATHarris Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl3-2 Resource RidsNrrDirsltsb Resource RidsNrrPMQuadCities Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource ccesslon ADAMS P ac k aqe A . NNRR 058 OFFICE LPL3-2/PM LPL3-2/LA DIRSIITSB OGC(NLO) LPL3-2/BC NAME CGratton THarris RElliot MSmith MDavid for SCampbeli DATE 10/9/09 10/9/09 9 I 25 109 10/16/09 10/21/09 OFFICIAL RECORD}}
: 2. Amendment No.           to DPR-30
: 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC                         RidsOgcRp Resource             LPL3-2 RlF RidsNrrLATHarris Resource     RidsNrrDorlDpr Resource       RidsNrrDorlLpl3-2 Resource RidsNrrDirsltsb Resource       RidsNrrPMQuadCities Resource RidsRgn3MailCenter Resource   RidsAcrsAcnw_MailCTR Resource ADAMS Packaqe A ccesslon  . No. ML092670488                                  NRR  058 OFFICE       LPL3-2/PM       LPL3-2/LA         DIRSIITSB       OGC(NLO)         LPL3-2/BC NAME         CGratton       THarris           RElliot         MSmith           MDavid for SCampbeli DATE           10/9/09       10/9/09           9 I 25 109       10/16/09         10/21/09 OFFICIAL RECORD COPY}}

Latest revision as of 08:28, 12 March 2020

Issuance of Amendments Risk Informed Modification to Selected Required Action End States for BWR Plants
ML092670488
Person / Time
Site: Quad Cities  Constellation icon.png
Issue date: 10/21/2009
From: Gratton C
Plant Licensing Branch III
To: Pardee C
Exelon Nuclear
Gratton C, NRR, DORL, 415-1055
References
TAC MD6997, TAC MD6998
Download: ML092670488 (56)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 21, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon l\Iuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RISK-INFORIVIED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AI\ID MD6998)

Dear Mr. Pardee:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession 1\10. ML090350151).

The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2."

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~~

Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket l\Ios. 50-254 and 50-265

Enclosures:

1. Amendment No. 245 to DPR-29
2. Amendment No. 240 to DPR-30
3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 245 Renewed License No. DPR-29

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC, et al.

(the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-29 is hereby amended to read as follows:

-2 B. Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 245 . are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~CfJJJ Stepte~ Campbell. Chief ro~

Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: October 21, 2009

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 Renewed License No. DPR-30

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC, et al. (the licensee) dated October 9,2007, as supplemented on January 30,2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the pUblic, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-30 is hereby amended to read as follows:

-2 B. Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 240

  • are hereby incorporated into the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/)v l{kj~~~jj ~((

Stephen Campbell, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: October 21, 2009

ATTACHMENT TO LICENSE AMENDMENT NOS. 245 AND 240 RENEWED FACILITY OPERATING LICENSES NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254 AND 50-265 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by number and contain marginal lines indicating the areas of change.

Remove License DPR-29 License DPR-29 Page 4 Page 4 License DPR-30 License DPR-30 Page 4 Page 4 TSs TSs 3.4.3-1 3.4.3-1 3.5.1-1 3.5.1-1 3.5.1-2 3.5.1-2 3.5.1-3 3.5.1-3 3.5.3-1 3.5.3-1 3.6.1.6-1 3.6.1.6-1 3.6.1.7-2 3.6.1.7-2 3.6.1.8-1 3.6.1.8-1 3.6.2.3-1 3.6.2.3-1 3.6.2.4-1 3.6.2.4-1 3.6.4.1-1 3.6.4.1-1 3.6.4.3-1 3.6.4.3-1 3.6.4.3-2 3.6.4.3-2 3.7.1-2 3.7.1-2 3.7.4-1 3.7.4-1 3.7.4-2 3.7.4-2 3.7.5-1 3.7.5-1 3.7.6-1 3.7.6-1 3.8.1-5 3.8.1-5 3.8.4-3 3.8.4-3 3.8.7-2 3.8.7-2

-4 B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.245 , are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Oder without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plans', which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2," submitted by letter dated May 17, 2006.

F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated july 27, 1979 with supplements dated November 5, 1980, and 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-29 Amendment No. 245

-4 B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No,- 24Q are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. The license shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D. Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1 , which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

"Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2,"

submitted by letter dated May 17, 2006.

F. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports dated .July 27 , 1979 with supplements dated 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-30 Amendment No. 240

Safety and Relief Valves 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety and Relief Valves LCO 3.4.3 The safety function of 9 safety valves shall be OPERABLE.

The relief function of 5 relief valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One relief valve A.1 Restore the relief 14 days inoperable. valve to OPERABLE status.

B. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met. ----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Two or more relief C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> valves inoperable.

AND OR C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> One or more safety valves inoperable.

Quad Cities 1 and 2 3.4.3-1 Amendment No245/240

ECCS-Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS-Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure ~ 150 psig.

ACTI ONS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQU I RED ACTI ON COMPLETION TIME A. One Low Pressure A.1 Restore LPCI pump to 30 days Coolant Injection OPERABLE status.

(LPCI) pump inoperable.

B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition subsystem to OPERABLE A. status.

OR One Core Spray subsystem inoperable.

C. One LPCI pump in each C.1 Restore one LPCI pump 7 days subsystem inoperable. to OPERABLE status.

D. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.

B, or C not met.

0.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (contlnued)

Quad Cities 1 and 2 3.5.1-1 Amendment No. 245/240

ECCS-Operating 3.5.1 ACTIONS CONDITION REQU I RED ACTI ON COMPLETION TIME E. Two LPCI subsystems E.l Restore one LPCI 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable for reasons subsystem to OPERABLE other than Condition status.

C.

F. Required Action and F.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition E AND not met.

F.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> G. HPCI System G.l Verify by Immediately inoperable. administrative means RCIC System is OPERABLE.

AND G.2 Restore HPCI System 14 days to OPERABLE status.

H. One ADS valve H.l Restore ADS valve to 14 days inoperable. OPERABLE status.

I. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition G or when entering MODE 3.

H not met.

1.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> J. Two or more ADS valves J.l Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

J.2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> dome pressure to S 150 psig.

(contlnued)

Quad Cities 1 and 2 3.5.1-2 Amendment No .245/240

ECCS-Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME K. Two or more low K.1 Enter LCO 3.0.3 Immediately pressure ECCS injection/spray subsystems inoperable for reasons other than Condition C or E.

HPCI System and one or more ADS valves inoperable.

One or more low pressure ECCS injection/spray subsystems inoperable and one or more ADS valves inoperable.

HPCI System inoperable and either one low pressure ECCS injection/spray subsystem is inoperable or Condition Centered.

Quad Cities 1 and 2 3.5.1-3 Amendment No.245/240

RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABI LITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.

ACTI ONS

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOT E- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by Immediately inoperable. administrative means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System 14 days to OPERABLE status.

B. Required Action and - - - - - - - - - - - - NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.5.3-1 Amendment No.245/240

Low Set Relief Valves 3.6.1.6 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low Set Relief Valves LCO 3.6.1.6 The low set relief function of two relief valves shall be OPERABLE.

APPLICABI LITY: MODES 1, 2, and 3.

ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME A. One low set relief A.1 Restore low set 14 days valve inoperable. relief valve to OPERABLE status.

B. Required Action and - - - - - - - - - - - - -NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met. ----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Two low set relief C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> valves inoperable.

AND C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Quad Cities 1 and 2 3.6.1.6-1 Amendment No. 245/240

Reactor BUilding-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

Associated Completion LCO 3.0.4.a is not applicable Time of Condition C not when entering MODE 3.

met. ----------------------------

0.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Two lines with one or E.1 Restore all vacuum 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> more reactor building- breakers in one line to-suppression chamber to OPERABLE status.

vacuum breakers inoperable for opening.

F. Required Action and F.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Associated Completion Time of Conditions A, AND B or E not met.

F.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.6.1.7.1 - - - - - - - - - - - - - - - - - - NOTES - - - - - - - - - - - - - - - - -

1. Not required to be met for vacuum breakers that are open during Surveillances.
2. Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed. 14 days SR 3.6.1.7.2 Perform a functional test of each vacuum 92 days breaker.

(continued)

Quad Cities 1 and 2 3.6.1.7-2 Amendment No. 245/240

Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6 CONTAINMENT SYSTEMS 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers LCO 3.6.1.8 Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening.

Twelve suppression chamber-to-drywell vacuum breakers shall be closed.

APPLICABILITY: MODES 1, 2, and 3.

ACTI ONS CONDITION REQUIRED ACT ION COMPLETION TIME A. One required A.1 Restore one vacuum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> suppression chamber breaker to OPERABLE to-drywell vacuum status.

breaker inoperable for opening.

B. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Conditi on A when entering MODE 3.

not met. ----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. One suppression C.1 Close the open vacuum 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> chamber-to-drywell breaker.

vacuum breaker not closed.

D. Required Action and 0.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C AND not met.

0.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Quad Cities 1 and 2 3.6.1.8-1 Amendment No. 245/240

RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTI ONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days poo 1 cooling subsY$tem suppression pool inoperable. cooling subsystem to OPERABLE status.

B. Required Action and - - - - - - - - - - - - - - NOT E- - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met. ----------------------------

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Two RHR suppression C.1 Restore one RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> poo 1 cooling suppression pool subsystems inoperable. cooling subsystem to OPERABLE status.

D. Required Action and 0.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C AND not met.

0.2 Be in I~ODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Quad Cities 1 and 2 3.6.2.3-1 Amendment No. 245/240

RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residua 1 Heat Remova 1 (RHR) Suppres s ion Pool Spray LCO 3.6.2.4 Two RHR suppression pool spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days pool spray subsystem suppression pool inoperable. spray subsystem to OPERABLE status.

B. Two RHR suppression B.1 Restore one RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> pool spray subsystems suppression pool inoperable. spray subsystem to OPERABLE status.

C. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.

C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.6.2.4-1 Amendment No. 245/240

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS COND ITION REQUI RED ACTION COMPLETION TIME A. Secondary containment A .1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, containment to 2, or 3. OPERABLE status.

B. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Secondary containment C.1 - - - - - - - - NOTE - - - - - - -

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.1-1 Amendment No. 245/240

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCD 3.6.4.3 Two SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT 7 days inoperable. subsystem to OPERABLE status.

B. Required Action and - - - - - - - - - - - - - - NOT E- - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Required Action and - - - - - - - - - - - -NOTE - - - - - - - - - -

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently C.1 Place OPERABLE SGT Immediately i rradi ated fuel subsystem in assemblies in the operation.

secondary containment or during OPDRVs.

(continued)

Quad Cities 1 and 2 3.6.4.3-1 Amendment No. 245/240

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

C.2.2 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Restore one SGT 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable in MODE 1, subsystem to 2, or 3. OPERABLE status.

E. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition D when entering MODE 3.

not met.

E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> F. Two SGT subsystems F.1 - - - - - - - -NOTE - - - - - -

inoperable during LCO 3.0.3 is not movement of recently applicable.

i rradi ated fuel assemblies in the secondary containment Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in secondary containment.

F.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.6.4.3-2 Amendment No. 245/240

RHRSW System 3.7.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and - - - - - - - - - - - - - - NOT E- - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Conditions A, when entering MODE 3.

B, or C not met.

D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Both RHRSW subsystems E.1 - - - - - - - - NOTE - - - - - - -

inoperable for reasons Enter applicable other than Conditions and Condition B. Required Actions of LCO 3.4.7 for RHR shutdown cooling subsystems made inoperable by RHRSW System.

Restore one RHRSW 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystem to OPERABLE status.

F. Required Action and F.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition E AND not met. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual and power operated 31 days valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

Quad Cities 1 and 2 3.7.1-2 Amendment No. 245/240

CREV System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Ventilation (CREV) System LCO 3.7.4 The CREV System shall be OPERABLE.

- - - - - - - - - - - - - - NOT E- - - - - - - - - - - - - - - - - -

The main control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTI ONS CONDITION REQUI RED ACTION COMPLETION TIME A. CREV System inoperable A.1 Restore CREV System 7 days in MODE 1, 2, or 3 for to OPERABLE status.

reasons other than Condition C.

B. Required Action and - - - - - - - - - - - - - - NOT E- - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. CREV system C.1 Initiate action to Immediately inoperable due to implement mitigating inoperable CRE actions.

boundary in MODE 1, 2, or 3.

C.2 Verify mitigating 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> actions ensure CRE occupant exposures to radiological, chemi cal, and smoke hazards wi 11 not exceed 1 i mits (continued)

Quad Cities 1 and 2 3.7.4-1 Amendment No. 245/240

CREV System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore CRE boundary 90 days to OPERABLE status D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C not met in MODE 1, 2, or 3. 0.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. CREV System inoperable - - - - - - - - - - - -NOTE - - - - - - - - - - - -

during movement of LCO 3.0.3 is not applicable.

recently irradiated fuel assembl i es in the secondary containment E.1 Suspend movement of Immediately or during OPDRVs. recently irradiated fuel assemblies in the secondary containment.

CREV System inoperable due to an inoperable AND CRE boundary during movement of recently E.2 Initiate action to Immediately irradiated fuel suspend OPDRVs.

assemblies in the secondary containment or during OPDRVs.

Quad Cities 1 and 2 3.7.4-2 Amendment No. 245/240

Control Room Emergency Ventilation AC System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Control Room Emergency Ventilation Air Conditioning CAC) System LCO 3.7.5 The Control Room Emergency Ventilation AC System shall be OPERABLE.

APPLICABI LITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel COPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Control Room Emergency A.1 Restore Control Room 30 days Ventilation AC System Emergency Ventilation inoperable in MODE 1, AC System to OPERABLE 2, or 3. status.

B. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.

not met in MODE 1, 2, or 3.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Control Room Emergency - - - - - - - - - - - - NOT E- - - - - - - - - - - -

Ventilation AC System LCO 3.0.3 is not applicable.

inoperable during movement of recently irradiated fuel C.1 Suspend movement of Immediately assemblies in the recently irradiated secondary containment fuel assemblies in or during OPDRVs. the secondary containment.

C.2 Initiate action to Immediately suspend OPDRVs.

Quad Cities 1 and 2 3.7.5-1 Amendment No. 245/240

Main Condenser Offgas 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Main Condenser Offgas LCO 3.7.6 The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be

~ 251,100 ~Ci/second after decay of 30 minutes.

APPLICA8I LITY: l"lODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Gross gamma activity A.1 Restore gross gamma 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> rate of the noble activity rate of the gases not withi n noble gases to withi n 1 i mi t. 1 i mi t.

8. Requi red Action and 8.1 Isolate all main 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion steam lines.

Time not met.

OR 8.2 Isolate SJAE. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR

- - - - - - - - - - - - - - NOTE - - - - - - - - - -

LCO 3.0.4.a is not applicable when entering MODE 3.

8.3 8e in l"lODE 3 . 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.7.6-1 Amendment No. 245/240

AC Sources-Operating 3.8.1 ACTIONS cCONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.

B, C, 0, or E not met. ----------------------------

F.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> G. Three or more required G.1 Enter LCO 3.0.3. Immediately AC sources inoperable.

Quad Cities 1 and 2 3.8.1-5 Amendment No. 245/240

DC Sources-Operating 3.8.4 ACTIONS CONDITION REQUI RED ACTION COMPLETION TIME D. Division 1 or 2 D.1 Restore Division lor 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 125 VDC electrical 2 125 VDC electrical power subsystem power subsystem to inoperable for reasons OPERABLE status.

other than Conditions B or C. OR D.2 --------NOTE--------

Only applicable if the opposite unit is not in MODE 1,2, or 3.

Place associated 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE alternate 125 VDC electrical power subsystem in service.

E. Opposite unit 125 VDC E.1 Restore the opposite 7 days electrical power unit 125 VDC subsystem inoperable. electrical power subsystem to OPERABLE status.

F. Required Action and - - - - - - - - - - - - - -NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Ti me not met. when entering MODE 3.

F.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Quad Cities 1 and 2 3.8.4-3 Amendment No. 245/240

Distribution Systems-Operating 3.8.7 ACTIONS CONDITION REQUI RED ACTION COMPLETION TIME B. One or more DC B.1 Restore DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> electrical power power distribution distribution subsystems to AND subsystems inoperable. OPERABLE status.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery of failure to meet LCO 3.8.7.a C. One or more required - - - - - - - - - - - - - NOTE - - - - - - - - - - -

opposite unit AC or DC Enter applicable Condition electrical power and Required Actions of distribution LCO 3.8.1 when Condition C subsystems inoperable. results in the inoperability of a required offsite circuit.

C.1 Restore required 7 days opposite unit AC and DC electrical power distribution subsystems to OPERABLE status.

D. Required Action and - - - - - - - - - - - - - - NOTE - - - - - - - - - -

associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.

B, or C not met.

D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Two or more electrical E.1 Enter LCO 3.0.3. Immediately power distribution subsystems inoperable that, in combination, result in a loss of function.

Quad Cities 1 and 2 3.8.7-2 Amendment No. 245/240

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDIVlEI\JT NO.245 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 EXELON GENERATION COMPANY, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265

1.0 INTRODUCTION

By letter dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072830096), as supplemented by letter dated January 30, 2009, (ADAMS Accession No. ML090350151), Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request (LAR) which proposed changes to the technical specifications (TSs) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The January 30, 2009, supplement contained clarifying information and did not change the NRC staff's initial proposed finding of no significant hazards consideration. The LAR would modify the TSs to risk-informed requirements regarding required action end states. In the request, QCNPS planned to adopt Technical Specification Task Force (TSTF) Change Traveler 423, (Reference (Ref.) 8), to the Boiling-Water Reactor (BWR) Standard Technical Specifications (STS) (NUREG 1433 and NUREG 1434), which was proposed by the TSTF Owners Group on August 12, 2003, on behalf of the industry. TSTF-423 incorporates the BWR Owners Group (BWROG) approved Topical Report (TR) NEDC-32988-A, Revision 2, "Technical Justification to Support Risk Informed Modification to Selected Required Action End States for BWR Plants" (BWROG TR, or Ref. 1), into the BWR STS (NOTE: The changes in TSTF-423 are made with respect to Revision 3 of the BWR STS NUREGs).

On March 30, 2001 (ADAMS Accession No ML011130309), the Nuclear Regulatory Commission (NRC, the Commission) staff approved the licensee's request to convert the QCNPS TSs to the improved TSs design based on NUREG-1433, Revision 1, "Standard Technical Specifications, General Electric Plants BWR/4," dated April 1995, and on guidance provided in the Commission's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published on July 22,1993 (58 FR 39132). The licensee's October 9,2007, application states that QCNPS TSs are based on NUREG-1433 though it is not identical to the Commission's Policy Statement guidance. Therefore, an adaptation of the referenced document was required.

-2 TSTF-423 is one of the industry's initiatives developed under the Risk Management Technical Specifications program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and risk management techniques in TSs, while reducing unnecessary burden and making TS requirement~ consistent with the Commission's other risk informed regulatory requirements, in particular, Title 10 of the Code of Federal Regulations (10 CFR), Section 50.65 (Ref. 3), the "Maintenance Rule." Section 50.36(c)(2)(i) of 10 CFR, states, in part: "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow the remedial action permitted by the technical specification until the condition can be met." Plant TSs provide, as part of the remedial action, a completion time (CT) for the plant to either comply with remedial actions or restore compliance with the limiting conditions for operation (LCD). If the LCD or the remedial action cannot be met, then the reactor is required to be shutdown. When the STS and individual plant TSs were written, the shutdown condition, or end state specified, was usually cold shutdown. The BWROG TR provides the technical basis to change certain required end states when the TS Actions for remaining in power operation cannot be met within the CTs. Most of the requested TS changes permit an end state of hot shutdown (Mode 3), if risk is assessed and managed, rather than an end state of cold shutdown (Mode 4), contained in the current TSs. The proposed changes were limited to those end states where: (1) entry into the shutdown mode is for a short interval, (2) entry is initiated by inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable TS, and (3) the primary purpose is to correct the initiating condition and return to power operation as soon as is practical.

The TSs for QCNPS define five operational modes:

In general, they are:

  • Mode 1 - Power Operation. The reactor mode switch is in run position.
  • Mode 2 - Reactor Startup. The reactor mode switch is in refuel position (with all reactor vessel head closure bolts fully tensioned) or in startup/hot standby position.
  • Mode 3 - Hot Shutdown. The reactor coolant system (RCS) temperature is above 212 degrees F (TS specific) and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tenSioned).
  • Mode 4 - Cold Shutdown. The RCS temperature is equal to or less than 212 degrees F and the reactor mode switch is in shutdown position (with all reactor vessel head closure bolts fully tensioned).
  • Mode 5 - Refueling. The reactor mode switch is in shutdown or refuel position, and one or more reactor vessel head closure bolts are less than fully tensioned.

Modifying the QCNPS TSs consistent with TSTF-423 allows a Mode 3 end state rather than a Mode 4 end state for selected initiating conditions in order to perform short-duration repairs.

Short duration repairs are on the order of 2-to-3 days, but not more than a week.

The licensee stated in its application that the BWROG TR and TSTF-423, as well as the NRC staff's safety evaluation (SE) (Ref. 6), were applicable to the QCNPS units, and provided justification for incorporation of the proposed changes into the QCNPS Units 1 and 2 TSs.

- 3

2.0 REGULATORY EVALUATION

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36(c), TSs are required to include items in the following eight specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.

The !\IRC staff did not review the LAR with respect to decommissioning, initial notification and written reports, as the licensee did not propose any changes to these specific requirements.

The rule does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(2)(i), the LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications. In describing the basis for changing end states, the BWROG TR states:

"Cold shutdown is normally required when an inoperable system or train cannot be restored to an operable status within the allowed time. Going to cold shutdown results in the loss of high pressure core cooling systems, challenges the shutdown heat removal systems, and requires restarting the plant. A more preferred operational mode is one that maintains adequate risk levels while repairs are completed without causing unnecessary challenges to plant equipment during shutdown and startup transitions."

In the end state changes under consideration, a problem with a component or train has, or will, result in a failure to meet TSs, and a controlled shutdown is directed because a TS Action statement cannot be met within the TS CT.

Most of today's TSs and the design basis analyses were developed under the perception that putting a plant in cold shutdown would result in the safest condition, and the design basis analyses would bound credible shutdown accidents. In the late 1980s and early 1990s, the NRC staff and licensees recognized that this perception was incorrect and took corrective actions to improve shutdown operation. At the same time, the STS were developed and many licensees took action to improve their TSs. Since enactment of a shutdown rule was expected, almost all TS changes involving power operation, including a revised end state requirement, were postponed (Ref. 2). However, in the mid-1990s, the Commission decided a shutdown rule was not necessary in light of industry improvements. Controlling shutdown risk encompasses control of conditions that can cause potential initiating events and responses to those initiating events that do occur. Initiating events are a function of equipment malfunctions and human error.

Responses to events are a function of plant sensitivity, ongoing activities, human error, defense in-depth, and additional equipment malfunctions.

In practice, the risk during shutdown operations is often addressed via voluntary actions and application of the Maintenance Rule (Ref. 3). Section 50.65(a)(4) of the Maintenance Rule states, in part:

"Before performing maintenance activities ... , the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of

- 4 the assessment may be limited to structures, systems, and components that a risk informed evaluation process has shown to be significant to public health and safety."

Regulatory Guide (RG) 1.182 (Ref. 4) provides guidance on implementing the provisions of 10 CFR 50.65(a)(4) by endorsing the revised Section 11 (published separately) to NUMARC 93-01, Revision 2 (Ref.7). The remainder of NUMARC 93-01, Revision 2, was previously endorsed by the NRC staff in RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," March 1997.

3.0 TECHNICAL EVALUATION

The changes proposed in the amendment are consistent with the changes proposed and justified in the BWROG TR and accepted by the !\IRC staff, as documented in the NRC staff's TR SE. The evaluation included in the NRC staffs TR SE, as appropriate and applicable to the changes of TSTF-423, is reiterated here, and differences from the SE are justified. In its application, the licensee commits to the implementation guidance for TSTF-423 contained in TSTF-IG-05-02 (Ref. 9), which addresses a variety of issues, such as considerations and compensatory actions for risk-significant plant configurations. An overview of the generic evaluation and associated risk assessment is provided below, along with a summary of the associated TS changes justified by the BWROG TR.

3.1 Risk Assessment The objective of the BWROG TR risk assessment was to show that any risk increases associated with the proposed changes in TS end states are either negligible or negative (Le., a net decrease in risk). The BWROG TR documents a risk-informed analysis of the proposed TS change. Probabilistic Risk Assessment (PRA) results and insights are used, in combination with results of deterministic assessments, to identify and propose changes in "end states" for all BWR plants. This is in accordance with guidance provided in RG 1.174 (Ref. 10) and RG 1.177 (Ref. 5). The three-tiered approach documented in RG 1.177 was followed. The first tier of the three-tiered approach includes the assessment of the risk impact of the proposed change for comparison to acceptance guidelines consistent with the Commission's Safety Goal Policy Statement, as documented in RG 1.174. The first tier aims at ensuring that there are no unacceptable temporary risk increases as a result of the TS change, such as when equipment is taken out of service. The second tier addresses the need to preclude potentially high-risk configurations which could result if equipment is taken out of service concurrently with the equipment out of service as allowed by this TS change. The third tier addresses the application of 10 CFR 50.65 (a)(4) of the Maintenance Rule for identifying risk-significant configurations resulting from maintenance-related activities and taking appropriate compensatory measures to avoid such configurations.

The proposed TS change invokes a risk assessment because 10 CFR 50.65(a)(4) is applicable to maintenance-related activities and does not cover other operational activities beyond the effect they may have on existing maintenance-related risk.

As discussed in the NRC staff's TR SE, the !\IRC staff found that the BWROG's risk assessment approach used in the BWROG TR was comprehensive and acceptable. In addition, the analyses show that the three-tiered approach criteria for allowing TS changes are met as follows:

-5

  • Risk Impact of the Proposed Change (Tier 1): The risk changes associated with the TS changes in TSTF-423, in terms of mean yearly increases in core damage frequency (CDF) and large early release frequency (LERF), are risk neutral or risk beneficial. In addition, there are no significant temporary risk increases, as defined by RG 1.177 criteria, associated with the implementation of the TS end state changes.
  • Avoidance of Risk-Significant Configurations (Tier 2): The risk analyses that were performed, which are based on single LCOs, indicate that there are no high-risk configurations associated with the TS end state changes. The reliability of redundant trains is normally covered by a single LCO. When multiple LCOs occur, which affect trains in several systems, the plant's risk-informed configuration risk management program, or the risk assessment and management program implemented in response to 10 CFR 50.65 (a)(4), shall ensure that high-risk configurations are avoided. As part of the implementation of TSTF-423, the licensee has committed to follow Section 11 of NUMARC 93-01, Revision 3, and include guidance in appropriate plant procedures and/or administrative controls to preclude high-risk plant configurations when the plant is at the proposed end state. While the NRC staff has not endorsed Revision 3 to NUMARC 93-01, the NRC staff has endorsed a revised version of NUMARC 93-01, Revision 2, Section 11 in RG 1.182. The !\IRC staff finds that such guidance is adequate for preventing risk-significant plant configurations.
  • Configuration Risk Management (Tier 3): The licensee has a commitment in place (as described below), to comply with 10 CFR 50.65 (a)(4) to assess and manage the risk from maintenance activities. This program can support the licensee's decision in selecting the appropriate actions to control risk for most cases in which a risk-informed TS is entered.

The generic risk impact of the end state mode change was evaluated subject to the following assumptions which are incorporated into the TS, TS Bases, and TSTF-IG-05-02:

a. The entry into the end state is initiated by the inoperability of a single train of equipment or a restriction on a plant operational parameter, unless otherwise stated in the applicable technical specification.
b. The primary purpose for entering the end state is to correct the initiating condition and return to power as soon as is practical.
c. When Mode 3 is entered as the repair end state, the time the reactor coolant pressure is above 500 psig will be minimized. If reactor coolant pressure is above 500 psig for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the associated plant risk will be assessed and managed.

These assumptions are consistent with typical entries into Mode 3 for short duration repairs, which is the intended use of the TS end state changes. The NRC staff concluded in its TR SE that, in general, going to Mode 3 instead of going to Mode 4 to carry out equipment repairs that are of short duration does not have any adverse effect on plant risk.

In its application, the licensee committed to follow the guidance established in Section 11 of NUMARC 93-01. NUMARC 93-01 provides guidance on implementing the provisions of 10 CFR 50.65(a)(4). The licensee also committed in the January 30,2009, supplement to follow the guidance established in TSTF-IG-05-02. The commitments are restated below:

-6 COMMITMENT TYPE COMMITMENT COI\/IMITTED ONE TIME PROGRAMMATIC DATE ACTION (YES/NO)

(YES/NO)

EGC will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Ongoing No Yes Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 3, July, 2000.

EGC will follow the guidance established in TSTF-IG-05-02, "Implementation Guidance for TSTF 423, Revision 0, 'Technical Specifications End States,NEDC 32988-A,'" Revision 1, March 2007.

Implement No Yes The follOWing statement on Page 2 with no longer applies: amendment "If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3."

The NRC staff notes that it has not endorsed NUMARC 93-01, Revision 3, referenced in the first commitment, but has evaluated the language in Section 11 of NUMARC 93-01, Revision 3, and finds that it is consistent with the version of Section 11 that has been endorsed by the NRC staff in RG 1.182. The NRC staff acceptance of this commitment relates only to Section 11 of NUMARC 93-01, Revision 3.

By following the implementation guidance, the licensee will ensure that defense-in-depth is maintained for key safety functions by ensuring availability of Tier 2 systems/equipment necessary for safe shutdown. Therefore, the NRC staff finds the licensee's commitments to be acceptable.

3.2 Request for Additional Information During its review of the application, the staff identified two concerns. First, revising the TS to allow the licensee to remain in Mode 3 indefinitely with inoperable systems would also permit starting up using the allowance of LCO 3.0.4(a) with inoperable systems or equipment. This is inconsistent with the purpose of TSTF-423, which is to allow licensees to remain in Mode 3 (instead of proceeding to Mode 4) while conducting repairs, and then return to Mode 1.

- 7 The second concern is that primary containment should not have been treated the same as the other systems included in the TSTF-423. Primary containment was not included in the TSTFs for the pressurized-water reactor designs (Le., TSTF-422 for Combustion Engineering Plants; TSTF-431 for Babcock and Wilcox plants). Unlike the other systems included in TSTF-423, an inoperable primary containment constitutes a loss of one of the three fission product barriers.

Staying at hot conditions in such an unanalyzed condition is not consistent with maintaining defense-in-depth, which is one of the five key principles of risk-informed regulations in RG 1.174.

From the RG perspective, the core damage risks are found to be acceptable; however, the compensatory measures identified (Le., availability of secondary containment, ventilation treatment systems, etc.) do not provide an acceptable defense-in-depth approach and, therefore, an equivalent level of protection, as provided by the primary containment, could not be attained by the compensatory measures.

To address these two concerns, the NRC staff issued a request for additional information (RAI)

(ADAMS Accession No. ML090080309) to the licensee on January 8, 2009. While the RAI was related to the TSTF-423 review for Clinton Nuclear Power Station, the NRC staff requested that the licensee's response include information for the three facilities with TSTF-423 applications under NRC staff review. The following summarizes the NRC staffs questions, and the licensee's responses for QCNPS:

RAI1: The licensee was requested to demonstrate how they would prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment.

RAI 2: The licensee was requested to demonstrate how they would maintain an equivalent level of protection while operating in Mode 3 with an inoperable primary containment.

On January 30, 2009, the licensee provided their response (ADAMS Accession No. ML090350151) to the NRC staff's RAI as follows:

Response to RA11: To prevent LCO 3.0.4(a) from being inappropriately invoked during startup to facilitate going up in mode with inoperable systems or equipment, EGC proposed the insertion of the following Note into those Required Actions affected by TSTF-423:

NOTE LCO 3.0.4.a is not applicable when entering MODE 3.

In addition, the licensee indicated that some of the previously submitted TS pages have been amended since its request to adopt TSTF-423 at QCNPS. Accordingly, the licensee provided revised versions of the TS pages that included the original TSTF-423 adoption markups and the above Note.

Response to RAI 2: The licensee indicated that it had evaluated its requests to amend station TS for primary containment and decided to withdraw its request to amend this TS. Because Mode 3 is no longer the requested end state for primary containment, the licensee determined that it is necessary to revise its original commitment to follow guidance established in TSTF-IG-05-02, to indicate that the following statement on Page 2 no longer applies:

"If Primary Containment is not operable, Secondary Containment and Standby Gas Treatment must be verified operable in order to remain in Mode 3."

- 8

==

Conclusion:==

The NRC staff reviewed the licensee's response to the staffs RAls, and found them to be acceptable since the amended station TSs prevent: a) operation in Mode 3 without primary containment, and b) starting up with inoperable systems or equipment.

3.3 Assessment of TS Changes The following sections discuss the specific changes, and include a synopsis of the STS LCOs.

The NRC staff discusses the acceptability of the proposed changes in Section 3.4.

3.3.1 LCO 3.4.3: Safety and Relief Valves The function of the safety valves is to protect the plant against severe overpressurization events.

The function of the relief valves is to control RCS pressure during transient conditions to prevent the need for safety valve actuation (except the S/RV: one of the safety valves also functions in relief mode) following such a transient.

LCO: The safety function of 9 safety valves and the relief function of 5 relief valves shall be OPERABLE.

Condition Requiring Entry into End State: If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the relief valve cannot be returned to operable status within that time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Modification for End State Required Actions: If the LCO cannot be met with one relief valve inoperable, the inoperable valve must be returned to operability within 14 days. If the valve cannot be returned to operable status within 14 days, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A Note is added to the TS Required Action for B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. If two or more relief valves or one or more safety valves become inoperable, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. The NRC staff reviewed the PRA evaluation and concluded in its TR SE that the core damage risks are approximately the same or lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 for one inoperable relief valve would cause loss of the High Pressure Coolant Injection (HPCI) system, Reactor Core Isolation Cooling (RCIC) system, and the power conversion systems due to the plant cooldown, and would require activating the Residual Heat Removal (RHR) system.

With one relief valve inoperable, the remaining valves are adequate to perform the required function. By remaining in Mode 3, HPCl j RCIC and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared to transitioning to Mode 4. In addition, the plant Emergency Operating Procedures (EOPs) direct the operators to take control of the depressurization function if low pressure injection/spray systems are needed for reactor pressure vessel (RPV) water makeup and cooling. The NRC staff concluded in its TR SE that the change allows repairs of the inoperable relief valve to be performed in a plant operating mode with lower risks.

- 9 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in the TS Required Action for B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with this system inoperable.

3.3.2 LCO 3.5.1: Emergency Core Cooling Systems (ECCS) - Operating The ECCS systems provide cooling water to the core in the event of a loss-of-coolant accident (LOCA). This set of ECCS TSs provides the operability requirements for the various ECCS subsystems as described below. This TS change would delete the secondary actions. The plant can remain in Mode 3 until the required repair actions are completed. The reactor is not depressurized.

LCO: Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.

Condition Requiring Entry into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions:

a. If one Low-Pressure Coolant Injection (LPCI) pump is inoperable, the LPCI pump must be restored to operable status in 30 days (Condition A).
b. If one LPCI subsystem is inoperable for reasons other than Condition A or one Core Spray subsystem is inoperable, the low pressure ECCS injection/spray subsystem must be restored to operable status within 7 days (Condition B).
c. If one LPCI pump in each subsystem is inoperable, one LPCI pump must be restored to operable status within 7 days (Condition C).
d. If two LPCI subsystems are inoperable for reasons other than Condition C, one LPCI 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 /> (Condition D).
e. If the Required Action and associated Completion Time of Condition A, B, C, or D is not met, then place the plant in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Condition E).
f. If the HPCI system is inoperable, verify immediately by administrative means that the RCIC system is operable and restore the HPCI system to operable status within 14 days (Condition F).
g. If one ADS valve is inoperable, it must be restored to operable status within 14 days (Condition G).
h. If the Required Action and associated Completion Time of Condition F or G is not met or two or more ADS valves become inoperable, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor steam dome pressure reduced to less than 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Condition H).
i. If two or more low pressure ECCS injection/spray subsystems are inoperable for reasons other than Condition C or D; or the HPCI system and one or more ADS

- 10 valves are inoperable; or one or more low pressure ECCS injection/spray subsystems are inoperable and one or more ADS valve are inoperable; or the HPCI system is inoperable and either one low pressure ECCS injection/spray subsystems is inoperable, or Condition C is entered, then LCO 3.0.3 must be entered immediately (Condition I).

Modification for End State Required Actions:

a. No change in Required Action for Condition A.
b. No change in Required Action for Condition B.
c. No change in Required Action for Condition C.
d. New Condition 0 states that, if the Required Action and associated Completion Time of Condition A, B, or C are not met, then place the plant in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (new Required Action 0.1). The plant is not taken into Mode 4 (cold shutdown). A Note is added to the TS Required Action 0.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Old Conditions 0, E, F and G are renumbered to E, F, G, and H with no changes to the Required Actions for these Conditions.
e. New Condition I states that, if the Required Action and associated Completion Time of renumbered Conditions G or H are not met, then the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (new Required Action I). The reactor is not depressurized and not taken to Mode 4. A Note is added to the TS Required Action 1.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.
f. Old Conditions H and I are renumbered to J and K. The renumbered Condition J states that if two or more ADS valves are inoperable, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor steam dome pressure reduced to less than 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. In renumbered Condition K, Condition 0 is changed to Condition E, in accordance with the reformatting of the "Conditions" column, with no changes to the Required Action and Completion Time for the Condition.

Assessment: The BWROG TR discusses a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state. The NRC staff's conclusion on the BWROG TR's PRA evaluation described in the NRC staffs TR SE indicates that the core damage risks are lower in Mode 3 than in the current end state of Mode 4. For QCNPS, going to Mode 4 for one ECCS subsystem or one ADS valve would cause loss of the HPCI, RCIC and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system. In addition, plant EOPs direct the operator to take control of the depressurization function if low-pressure injection/spray systems are needed for RPV water makeup and cooling.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions 0.1 and 1.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

- 11 3.3.3 LCD 3.5.3: Reactor Core Isolation Cooling System The function of the RCIC system is to provide reactor coolant makeup during loss of feedwater and other transient events. This TS provides the operability requirements for the RCIC system as described below. In the event of an inoperable RCIC system, the TS change allows the plant to remain in Mode 3 until the repairs are completed.

LCD: The RCIC system must be operable during Mode 1. During Modes 2 and 3, RCIC system must be operable when the reactor steam dome pressure is greater than 150 psig.

Condition Requiring Entry into End State: If the LCD cannot be met, the following actions must be taken: (a) verify immediately by administrative means that the HPCI system is operable (Required Action A.1), and (b) restore the RCIC system to operable status within 14 days (Required Action A.2). If either or both actions cannot be completed within the allotted time, the plant must be in placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor steam dome pressure reduced to less than 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Actions B.1 and B.2).

Modification for End State Required Actions: This TS change keeps the plant in Mode 3 until the required repairs are completed. A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. The reactor steam dome pressure is not reduced to less than 150 psig (delete Required Action B.2).

Assessment: This change would allow the inoperable RCIC system to be repaired in a plant operating mode with lower risk and without challenging the normal shutdown systems. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. In the NRC staff's TR SE, it concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 3 with reactor steam dome pressure less than 150 psig would cause loss of HPCI and a loss of the power conversion systems, and would require activating the RHR system. By remaining in Mode 3 above 150 psig steam dome pressure, HPCI and the power conversion systems remain available for coolant inventory control and decay heat removal. In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling.

The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.4 LCD 3.6.1.6: Low Set Relief Valves The function of low set relief valves is to prevent excessive short-duration relief valve cycling during an overpressure event. The TS provides operability requirements for the two low set relief valves as described below. The TS change allows the plant to remain in Mode 3 until the repairs are completed.

LCD: The low set relief function of two relief valves shall be OPERABLE.

Condition Requiring Entry into End State: If one low set relief valve is inoperable, it must be returned to operability within 14 days. If the low set relief valve cannot be returned to operable

- 12 status within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Modification for End State Required Actions: The TS change would allow the plant to stay in Mode 3 until the required repair actions are completed. The plant would not be taken into Mode 4 (cold shutdown) (delete Required Action B.2). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. (!'Jote: the Required Action for two low set relief valves inoperable was changed from Condition B to new Condition C without changing the Required Action end state.)

Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and the Mode 3 end state was performed. In the NRC staff's TR SE, the staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause loss of the HPCI, RCIC, and loss of the power conversion systems due to the plant cooldown, and would require activating the RHR system.

With one low set relief valve inoperable, the remaining valves are adequate to perform the required function. In addition, the plant EOPs direct the operator to take control of the depressurization function if low pressure injection/spray systems are needed for RPV water makeup and cooling. The NRC staff concluded in its TR SE that this change allows repairs of the inoperable low set relief valve to be performed in a plant operating mode with lower risks.

The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.

3.3.5 LCO 3.6.1.7: Reactor Building-to-Suppression Chamber Vacuum Breakers The reactor building-to-suppression chamber vacuum breakers relieve vacuum when the primary containment depressurizes below the pressure of the reactor building, thereby serving to preserve the integrity of the primary containment.

LCO: Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE.

Condition Requiring Entry into End State: If one line has one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening, the breaker(s) must be returned to operability within 7 days (Required Action C.1). If the vacuum breaker(s) cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action E.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action E.2).

Modification for End State Required Actions: A new Condition D modifies the Required Actions so that if the vacuum breaker(s) cannot be returned to operable status within the required Completion Times, the plant is placed in hot shutdown. A !'Jote is added to the TS Required Action D.1 stating that LCO 3.0A.a is not applicable when entering Mode 3. Existing Conditions D and E are renumbered to E and F without changing the Required Action end state. Condition F would require shutting down the plant to Mode 3 (Required Action F.1) and then Mode 4 (Required Action F.2), to address an inability to comply with the required actions related to Conditions A, B, and E.

Assessment: In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The NRC staff evaluated this conclusion in its TR SE. The

- 13 reduced end state would only be applicable to the situation where the vacuum breaker(s) in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function. The existing end state remains unchanged, as established by new Condition F, for conditions involving more than one inoperable line or vacuum breaker, since they are needed in Modes 1, 2, and 3. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling.

Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3. In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling.

The Note "LCD 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1 prevents an inappropriate use of the LCD 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.6 LCD 3.6.1.8: Suppression Chamber-to-Drywell Vacuum Breakers The function of the suppression chamber-to-drywell vacuum breakers is to relieve vacuum in the drywell, thereby preventing an excessive negative differential pressure across the wetwell/drywell boundary.

LCD: Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening, AND 12 suppression chamber-to-drywell vacuum breakers shall be closed.

Condition Requiring Entry into End State: If one suppression chamber-to-drywell vacuum breaker is inoperable for opening, the vacuum breaker must be returned to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Required Action A.1). If the vacuum breaker cannot be returned to operability within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action C.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action C.2).

Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if vacuum breaker(s) cannot be returned to operable status within the required Completion Time, the plant is placed in hot shutdown. A Note is added to the TS Required Action B.1 stating that LCD 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band C are renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) and Mode 4 (Required Action D.2), to address an inability to comply with the required actions related to Condition C.

Assessment: In the BWROG TR, it was determined that the specific failure condition of interest is not risk significant in BWR PRAs. The reduced end state would only be applicable to the situation where one or more vacuum breakers in one line are inoperable for opening, with the remaining operable vacuum breakers capable of providing the necessary vacuum relief function.

In the NRC staff's TR SE, the staff concluded that by remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, maintaining defense-in-depth. In addition, the plant EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for reactor coolant makeup and cooling.

- 14 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.7 LCO 3.6.2.3: RHR Suppression Pool Cooling Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by two redundant RHR suppression pool cooling subsystems.

LCO: Two RHR suppression pool cooling subsystems shall be OPERABLE.

Condition Requiring Entry into End State: If one RHR suppression pool cooling subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If two RHR suppression pool cooling subsystems are inoperable (Condition B), one RHR suppression pool cooling system 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 /> (Required Action B.1). If the RHR suppression pool cooling subsystem cannot be restored to operable status within the allotted time (Condition C), the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action C.1), and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action C.2).

Modification for End State Required Actions: A new Condition B modifies the Required Actions so that if the Required Action and associated Completion Time of Condition A are not met, the plant is placed in hot shutdown (Mode 3). A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. Existing Conditions Band Care renumbered to C and D without changing the Required Action end state. Condition D would require shutting down the plant to Mode 3 (Required Action D.1) after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and to Mode 4 (Required Action D.2) after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, to address an inability to comply with the required actions related to Condition C.

Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 end state was completed. The results described in the BWROG TR, and evaluated by the NRC staff in its TR SE, indicated that the core damage risks while operating in Mode 3 (assuming the individual failure conditions) are lower or comparable to the current end state. For QCNPS, one loop of the suppression pool cooling system is sufficient to accomplish the required safety function. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray is needed for RCS makeup and cooling.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.8 LCO 3.6.2.4: RHR Suppression Pool Spray Following a design-basis accident (DBA), the RHR suppression pool spray system removes heat from the suppression chamber airspace. A minimum of one RHR suppression pool spray subsystem is required to mitigate potential bypass leakage paths from the drywell and maintain the primary containment peak pressure below the design limits.

- 15 LCO: Two RHR suppression pool spray subsystems shall be OPERABLE.

Condition Requiring Entry into End State: If one RHR suppression pool spray subsystem is inoperable (Condition A), it must be restored to operable status within 7 days (Required Action A.1). If both RHR suppression pool spray subsystems are inoperable (Condition B), one of them 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 /> (Required Action B.1). If the RHR suppression pool spray subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action C.1), and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action C.2).

Modification for End State Required Actions: Required Action C.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action C.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: The main function of the RHR suppression spray system is to remove heat from the suppression chamber so that the pressure and temperature inside primary containment remain within analyzed design limits. The RHR suppression spray system was designed to mitigate potential effects of a postulated DBA, that is, a large LOCA which is assumed to occur concurrently with the most limiting single failure, and assuming conservative inputs. Under the conditions assumed in the DBA, steam blown down from the break could bypass the suppression pool and end up in the suppression chamber air space. The RHR suppression spray system could be needed to condense such steam so that the pressure and temperature inside primary containment remain within analyzed design limits.

However, the frequency of a DBA is very small and the containment has considerable margin to failure above the design limits. For these reasons, the unavailability of one or both RHR suppression spray subsystems has no significant impact on CDF or LERF, even for accidents initiated during operation at power. Therefore, it is very unlikely that the RHR suppression spray system will be challenged to mitigate an accident occurring during power operation. This probability becomes extremely unlikely for accidents that would occur during a small fraction of the year (less than 3 days) during which the plant would be in Mode 3 (associated with lower initial energy level and reduced decay heat load as compared to power operation) to repair the failed RHR suppression spray system.

Section 6 of the NRC staff's TR SE summarizes the staff's risk argument for approval of the BWROG TR's Section 4.5.1.11 and LCO 3.6.2.4. The argument for staying in Mode 3 instead of going to Mode 4 to repair the RHR Suppression Pool Spray system (one or both trains) is also supported by defense-in-depth considerations. TR SE Section 5.2 of the NRC staff's SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases, and precludes the need for RHR suppression pool spray subsystems. In addition, the probability of a DBA (large-break) is much smaller during shutdown as compared to power operation. A DBA in Mode 3 would be considerably less severe than a DBA occurring during power operation since Mode 3 is associated with lower initial energy level and reduced decay heat load. Under these extremely unlikely conditions, an alternate method that can be used to remove heat from the primary containment (in order to keep the pressure and temperature within the analyzed design basis limits) is containment venting. For more realistic accidents that could occur in Mode 3, several alternate means are available to remove heat from the primary containment, such as the RHR system in the suppression pool cooling

- 16 mode and the containment spray mode. The risk and defense-in-depth arguments used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable RHR suppression pool spray system.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.9 LCO 3.6.4.1: Secondary Containment Following a DBA, the function of the secondary containment is to contain, dilute, and stop radioactivity (mostly fission products) that may leak from primary containment. Its leak tightness is required to ensure that the release of radioactivity from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the standby gas treatment (SGT) system prior to discharge to the environment.

LCO: The secondary containment shall be OPERABLE.

Condition Requiring Entry into End State: If the secondary containment is inoperable, it must be restored to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (Required Action A.1). If it cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.1), and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action B.2).

Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the TS Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: This LCO entry condition does not include gross leakage through an unisolable release path. In the BWROG TR, it was concluded that previous generic PRA work related to 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," requirements has shown that containment leakage is not risk significant. The primary containment and all other primary and secondary containment-related functions would still be operable, including the SGT system, thereby minimizing the likelihood of an unacceptable release. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling, improving defense-in-depth compared with transitioning to Mode 4.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action Action B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.10 LCO 3.6.4.3: Standby Gas Treatment System The function of the SGT system is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a DBA are filtered and adsorbed prior to exhausting to the environment.

LCO: Two SGT subsystems shall be OPERABLE.

- 17 Condition Requiring Entrv into End State: If one SGT subsystem is inoperable, it must be restored to operable status within 7 days (Required Action A.1). If the SGT subsystem cannot be restored to operable status within the allotted time, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action B.2). In addition, if two SGT subsystems are inoperable in Mode 1, 2, or 3, (Condition D), and one SGT system can not be restored to operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> then the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action E.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action E.2).

Modification for End State Required Actions: Required Action B.2 and E.2 are deleted allowing the plant to stay in l\IIode 3 while completing repairs. A Note is added to the TS Required Actions B.1 and E.1 stating that LCO 3.0A.a is not applicable when entering Mode 3.

Assessment: The unavailability of one or both SGT subsystems has no impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the SGT system (i.e., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases resulting from materials that leak from the primary to the secondary containment above TS limits) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (i.e., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.

Section 5.2 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Sections 4.5.1.13 and 4.5.2.11, and LCO 3.6.4.3. According to this evaluation, which applies to BWR plant types, (such as, QCNPS's BWR-3 design), staying in Mode 3 instead of going to Mode 4 to repair the SGT system (one or both trains) is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable SGT system.

The Note "LCO 3.0A.a is not applicable when entering Mode 3" in TS Required Actions B.1 and E.1 prevents an inappropriate use of the LCO 3.0A.a allowance to go up in Mode with inoperable systems or equipment.

3.3.11 LCO 3.7.1: Residual Heat Removal Service Water (RHRSW) System The RHRSW system is designed to provide cooling water for the RHR system heat exchangers, which are required for safe shutdown following a normal shutdown or DBA or transient.

LCO: Two RHRSW subsystems shall be OPERABLE.

Condition Requiring Entrv into End State: If the LCO cannot be met, the following actions must be taken for the listed conditions:

a. If one RHRSW pump is inoperable (Condition A), it must be restored to operable status within 30 days (Required Action A.1).

- 18

b. If one RHRSW pump in each subsystem is inoperable (Condition B), one RHRSW pump must be restored to operable status within 7 days (Required Action B.1).
c. If one RHRSW subsystem is inoperable for reasons other than Condition A (Condition C), the RHRSW subsystem must be restored to operable status within 7 days (Required Action C.1).
d. If both RHRSW subsystems are inoperable for reasons other than Condition B (Condition D), then one RHRSW subsystem must be restored to operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
e. If the required action and associated completion time cannot be met within the allotted time (Condition E), the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action E.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action E.2)

Modification for End State Required Actions: A new Condition D is added which establishes requirements for existing Conditions A, B, and C, that are similar to existing Condition E, but without Required Action E.2. A Note is added to the Required Actions D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3. The existing Required Conditions D and E are renumbered to Conditions E and F, respectively. Required Condition F now applies to Condition E, which maintains the existing requirements with respect to both RHRSW subsystems being inoperable for reasons other than Condition B.

Assessment: In the BWROG TR, a comparative PRA evaluation of the core damage risks when operating in the current end state versus the Mode 3 end state was performed. The NRC staff conclusions, discussed in its TR SE, indicate that the core damage risks are lower when remaining in Mode 3 compared with transitioning to the current end state of Mode 4. By remaining in Mode 3, HPCI, RCIC, and the power conversion systems remain available to ensure adequate core cooling. Additionally, the EOPs direct the operators to take control of the depressurization function if low pressure injection/spray are needed for RCS makeup and cooling. Therefore, defense-in-depth is improved with respect to water makeup and decay heat removal by remaining in Mode 3, and the required safety function can still be performed with the RHRSW subsystem components that are still operable.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions D.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.12 LCO 3.7.4: Control Room Emergency Ventilation (CREV) system The CREV system provides a radiologically controlled environment from which the plant can be safely operated following a DBA.

LCO: The CREV system shall be OPERABLE.

Condition Requiring Entry into End State: If the CREV system is inoperable in Mode 1,2, or 3 for reasons other than an inoperable control room envelope (CRE) boundary in Mode 1, 2, or 3, (Condition B), it must be restored to operable status within 7 days (Required Action A.1). If the

- 19 CREV system is inoperable in Mode 1, 2, or 3 due to inoperable Control Room Envelope boundary, the licensee must initiate mitigating actions immediately (Required Action B.1), verify mitigating actions ensure CRE occupant exposures to airborne hazards will not exceed limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Required Actions B.2), and restore CRE operability within 90 days (Required Actions B.3). If the CREV system cannot be restored to operable status within the allotted time for Condition A or B, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action C.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action Co2).

Modification for End State Required Actions: The change adds a new Condition B with Required Action B.1 to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when Required Action and associated Completion Time of Condition A are not met in Mode 1, 2, or 3. The change renumbers old Conditions B, C and D to Conditions C, D and E. The renumbered Condition D would require that if the CREV system cannot be restored to operable status within the allotted time for renumbered Condition C, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action D.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action D.2). A Note is added to the new Required Actions B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: The unavailability of the CREV system has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the CREV system (Le., the frequency with which the system is expected to be challenged to provide a radiologically-controlled environment in the main control room following a DBA which leads to core damage and leaks of radiation from the containment that can reach the control room) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.

Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (i.e., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.16, and LCO 3.7.4, "Main Control Room Environmental Control (MCREC) System." (Note: The CREV system at QCNPS serves a similar design purpose as the MCREC described in NUREG 1433). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable CREV system.

The Note "LCO LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.13 LCO 3.7.5: Control Room Emergency Ventilation Air Conditioning System The Control Room AC system provides temperature control for the control room following control room isolation during accident conditions.

LCO: The Control Room Emergency Ventilation Air Conditioning System shall be OPERABLE.

- 20 Condition Requiring Entry into End State: If the Control Room Emergency Ventilation Air Conditioning System is inoperable, the system must be restored to operable status within 30 days (Required Action A.1). If the required actions and associated completion times for Condition A cannot be met, the plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action B.2).

Modification for End State Required Actions: Required Action B.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. A Note is added to the Required Action B.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: The unavailability of the Control Room Emergency Ventilation Air Conditioning System has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the system (Le., the frequency with which the system is expected to be challenged to provide temperature control for the control room following control room isolation following a DBA) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively), is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.

Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.17, and LCO 3.7.5, "Control Room Air Conditioning (AC) System," (which functions similar to QCNPS's Control Room Emergency Ventilation Air Conditioning System). The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the system is also supported by defense-in depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable Control Room Emergency Ventilation Air Conditioning System.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.14 LCO 3.7.6: Main Condenser Offgas The offgas from the main condenser normally includes radioactive gases. The gross gamma activity rate is controlled to ensure that accident analysis assumptions are satisfied and that offsite dose limits will not be exceeded during postulated accidents. The Main Condenser Offgas (MCOG) gross gamma activity rate is an initial condition of a DBA which assumes a gross failure of the MCOG system pressure boundary.

LCO: The gross gamma activity rate of the noble gases measured prior to the offgas holdup line shall be S 251,100 IJCi/second after decay of 30 minutes.

- 21 Condition Requirinq Entry into End State: If the gross gamma activity rate of the noble gases in the MCOG system is not within limits, the gross gamma activity rate of the noble gases in the main condenser Offgas must be restored to within limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Required Action A.1).

If the required action and associated completion time cannot be met, one of the following must occur:

a. All main steamlines must be isolated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.1).
b. The steam jet air ejector must be isolated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.2).
c. The plant must be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.3.1) and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action 8.3.2).

Modification for End State Required Actions: Required Action B.3.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. Required Action B.3.1 is renumbered to Required Action B.3 and a Note is added to the TS Required Action B.3 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: The failure to maintain the gross gamma activity rate of the noble gases in the MCOG within limits has no significant impact on CDF or LERF, irrespective of the mode of operation at the time of the accident. Furthermore, the challenge frequency of the MCOG system (Le., the frequency with which the system is expected to be challenged to mitigate offsite radiation releases following a DBA) is less than 1.0E-6/yr. Consequently, the conditional probability that this system will be challenged during the repair time interval while the plant is at either the current or the proposed end state (Le., Mode 4 or Mode 3, respectively) is less than 1.0E-8. This probability is considerably smaller than probabilities considered "negligible" in RG 1.177 for much higher consequence risks, such as large early release.

Section 5 of the NRC staff's TR SE makes a comparison between the Mode 3 and the Mode 4 end states, with respect to the means available to perform critical functions (Le., functions contributing to the defense-in-depth philosophy) whose success is needed to prevent core damage and containment failure and mitigate radiation releases. Section 6 of the NRC staff's TR SE summarizes its risk argument for approval of TR Section 4.5.1.18 and LCO 3.7.6. The argument for staying in Mode 3 instead of going to the Mode 4 end state to repair the MCOG system (one or both trains) is also supported by defense-in-depth considerations. The risk and defense-in-depth arguments, used according to the "integrated decision-making" process of RGs 1.174 and 1.177, support the conclusion that Mode 3 is as safe as Mode 4 for repairing an inoperable MCOG system.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Actions B.3.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.15 LCO 3.8.1: Alternating Current (AC) Sources - Operating The purpose of the AC electrical system is to provide the power required to put and maintain the plant in a safe condition and prevent the release of radioactivity to the environment. The Class 1E electrical power distribution system AC sources consist of the offsite power source (preferred power sources, normal and alternate(s)), and the onsite standby power sources (e.g., diesel generators (DGs)).

- 22 Per the QCNPS Updated Final Safety Analysis Report, the design of the AC electrical system provides independence and redundancy. The onsite Class 1E AC distribution system is divided into redundant divisions so that the loss of anyone division does not prevent the minimum safety functions from being performed. Each division has connections to two preferred offsite power sources and a single DG or other Class 1E Standby AC power source.

Offsite power at QCNPS is supplied to the unit switchyard(s) from the transmission network by five 345-kV transmission lines. Electric energy generated at the station is stepped up to 345-kV by the main power transformers and fed into the station's 345-kV transmission terminal ring bus which is also connected to the five transmission lines. The 345-kV ring bus provides power to two reserve auxiliary power transformers. Each reserve auxiliary transformer has sufficient capacity to handle the auxiliary power requirements of one unit. Each of these auxiliary power supplies is available, through circuit breaker switching, to both Division I and the Division II emergency auxiliary equipment of both units, and therefore, serves as a redundant offsite source of auxiliary power. In the event of a loss of offsite power, the emergency electrical loads are automatically connected to the operating DGs in sufficient time to provide for a safe reactor shutdown and to mitigate the consequences of a DBA.

LCO: The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class1 E AC Electric Power Distribution System.
b. Two DGs
c. One qualified circuit between the offsite transmission network and the opposite unit's onsite Class 1E AC Electrical Power Distribution System capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5. (Unit 2 only); and
d. The opposite unit's DG capable of supporting the equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), and LCO 3.7.5 (Unit 2 only).

Condition Requiring Entry into End State: Plant operators must bring the plant to 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 /> following the sustained inoperability of either or both required offsite circuits; one or two required DGs; or one required offsite circuit and one required DG.

Modification for End State Required Actions: Required Action F.2 is deleted allowing the plant to stay in Mode 3 while completing repairs. The plant will remain in Mode 3 (hot shutdown)

(Required Action F.1). A Note is added to the TS Required Action F.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: Entry into any of the conditions for the AC power sources implies that the AC power sources have been degraded and the single failure protection for the safe shutdown equipment may be ineffective. Consequently, as specified in the current STS 3.8.1, the plant operators must bring the plant to Mode 4 when the required action is not completed by the specified time for the associated action.

- 23 In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed. Events initiated by the loss of offsite power are dominant contributors to core damage frequency in most BWR PRAs, and RCIC and HPCI play major roles in mitigating these events. In the !\JRC staft's TR SE, the NRC staff concluded that, based on the low probability of loss of the AC power and the number of systems available in Mode 3, the core damage risks are lower in Mode 3 than in Mode 4 for one inoperable AC power source. For QCNPS, going to Mode 4 for one inoperable AC power source would cause loss of RCIC and HPCI, and loss of the power conversion systems, and would require activating the RHR system.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action for F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.16 LCO 3.8.4: Direct Current (DC) Sources - Operating The purpose of the DC power system is to provide a reliable source of DC power for both normal and abnormal conditions. It must supply power in an emergency for an adequate length of time until normal supplies can be restored.

LCO: For Modes 1, 2, and 3, the following DC electrical power subsystems shall be OPERABLE:

a. Two 250 VDC electrical power subsystems;
b. Division 1 and Division 2 125 VDC electrical power subsystems: and
c. The opposite unit's 125 VDC electrical power subsystem capable of supporting equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only),

and LCO 3.8.1.

Condition Requiring Entry into End State: The plant operators must bring the plant to Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action F.1) and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Required Action F.2) following the sustained inoperability of one 250 VDC electrical subsystem, or Division 1 or 2, DC electrical power subsystem for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Conditions A thru D), or 7 days if opposite unit 125 VDC electrical power subsystem is inoperable (Condition E).

Modification for End State Required Actions: Required Action F.2 is deleted, allowing the plant to stay in Mode 3 while Conditions A thru E are not met. A !\Jote is added to the TS Required Action F.1 stating that LCO 3.0.4(a) is not applicable when entering Mode 3.

Assessment: If one of the DC electrical power subsystems is inoperable, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the Mode 3 and Mode 4 end states was performed, with one DC system inoperable. Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the RCIC and HPCI systems play major roles in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4.

For QCNPS, going to Mode 4 would cause loss of the RCIC, HPCI and the power conversion systems (condenser/feedwater), and would require activating the RHR system.

- 24 The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action F.1 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.3.17 LCO 3.8.7: Distribution Systems - Operating The onsite Class 1E AC electrical power distribution system is divided into redundant and independent AC electrical power distribution systems. The primary AC electrical power distribution subsystem for each plant consists of two 4.16-kV Essential Service System (ESS) buses having an offsite source of power as well as a dedicated onsite DG source. The secondary plant distribution subsystems include 480-VAC ESS buses and associated load centers, motor control centers, distribution panels and transformers.

The 120-VAC vital buses are arranged in three different subsystems: 120V reactor protection system, 120V instrumentation bus system, and 120V ESS. The 120V ESS bus is supplied by a static uninterruptible power supply. There are two independent 250 VDC station service electrical power distribution subsystems and two independent 125 VDC electrical power distribution subsystems.

LCO: For Modes 1, 2, and 3, the following electrical power distribution subsystems shall be OPERABLE:

a. Division 1 and Division 2 AC and DC electrical power distribution subsystems; and
b. The portions of the opposite unit's AC and DC electrical power distribution subsystems necessary to support equipment required to be OPERABLE by LCO 3.6.4.3, LCO 3.7.4 (Unit 2 only), LCO 3.7.5 (Unit 2 only), and LCO 3.8.1.

Condition Requiring Entry into End State: The plant operators must bring the plant to 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 /> following the sustained inoperability of one or more AC, DC, or one or more required opposite unit AC or DC electrical power distribution subsystems inoperable for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 7 days, respectively (with a maximum 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Completion Time limit from initial discovery of failure to meet the LCO, to preclude being in the LCO indefinitely).

Modification for End State Required Actions: The TS change is to remove the requirement to place the plant in Mode 4 (Required Action D.2 is deleted). A Note is added to the TS Required Action D.1 stating that LCO 3.0.4.a is not applicable when entering Mode 3.

Assessment: If one of the AC/DC/AC ESS is inoperable, the remaining AC/DC/AC ESS have the capacity to support a safe shutdown and to mitigate an accident condition. In the BWROG TR, a comparative PRA evaluation of the core damage risks of operation in the current end state and in the Mode 3 end state was performed with one of the AC/DC/AC ESS inoperable. Events initiated by the loss of offsite power are dominant contributors to CDF in most BWR PRAs, and the high pressure core cooling systems, HPCI and RCIC, playa major role in mitigating these events. In the TR SE, the NRC staff concluded that the core damage risks are lower in Mode 3 than in Mode 4. For QCNPS, going to Mode 4 would cause a loss of the RCIC and HPCI systems, and the power conversion systems, and would require activating the RHR system.

The Note "LCO 3.0.4.a is not applicable when entering Mode 3" in TS Required Action D.1

- 25 prevents an inappropriate use of the LCO 3.0.4.a allowance to go up in Mode with inoperable systems or equipment.

3.4. Overall Assessment of Proposed Technical changes:

Based upon the above assessments, and because the time spent in Mode 3 to perform the repair on any of the systems described above would be infrequent and limited, and in light of defense-in-depth considerations (discussed above and in the BWROG TR, and as evaluated by the NRC staff's TR SE), the NRC staff concludes the changes to the QCNPS TSs described above are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that adopting TSTF-423, Rev 0, involves no significant hazards considerations, and there has been no public comment on the finding in Federal Register Notice 70 FR 74037, December 14, 2005. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, on the basis of the considerations discussed above, that (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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. NEDC-32988-A, Revision 2, "Technical Justification to Support Risk-Informed Modification to Selected Required Action End States for BWR Plants," December 2002.

(ADAMS Accession No. ML030170084).

2. Federal Register, Vol. 58, No. 139, p. 39136, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Plants," July 22,1993 (58 FR 39132).
3. 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."

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4. Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000 (ADAMS Accession No. ML003699426).
5. Regulatory Guide 1.177, "An Approach for Plant Specific, Risk-Informed Decision Making: Technical Specifications," USNRC, August 1998. (ADAMS Accession 1\10.

ML003740176).

6. NRC Safety Evaluation for Topical Report NEDC-32988, Revision 2, September 27, 2002. (ADAMS Accession No. ML022700603).
7. NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision 2, April 1996.
8. TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A." (ADAMS Accession No. ML032270250).
9. TSTF-IG-05-02, "Implementation Guidance for TSTF-423, Revision 0, 'Technical Specifications End States,' NEDC-32988-A," September 2005. (ADAMS Accession No. ML052700156).
10. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis,"

USNRC, August 1998. (ADAMS Accession No. ML003740133).

Principal Contributors: R. P. Grover, NRR Kristy Bucholtz, NRR Date: October 21, 2009

Mr. Charles G. Pardee October 21,2009 President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: RISK-INFORMED MODIFICATION TO SELECTED REQUIRED ACTION END STATES FOR BOILING-WATER REACTOR PLANTS (TAC NOS. MD6997 AND MD6998)

Dear Mr. Pardee:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 245 to Renewed Facility Operating License No. DPR-29 and Amendment No. 240 to Renewed Facility Operating License No. DPR-30 for the Quad Cities Nuclear Power Station, Units 1 and 2, respectively. The amendments are in response to your application dated October 9, 2007 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML072830096), as supplemented by letter dated January 30, 2009 (ADAMS Accession No. ML090350151).

The amendments would modify the technical specifications to risk-informed requirements regarding selected required action end states as provided in Technical Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 0, "Technical Specifications End States, NEDC-32988-A, Revision 2."

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Christopher Gratton, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosures:

1. Amendment No. to DPR-29
2. Amendment No. to DPR-30
3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsOgcRp Resource LPL3-2 RlF RidsNrrLATHarris Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl3-2 Resource RidsNrrDirsltsb Resource RidsNrrPMQuadCities Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource ADAMS Packaqe A ccesslon . No. ML092670488 NRR 058 OFFICE LPL3-2/PM LPL3-2/LA DIRSIITSB OGC(NLO) LPL3-2/BC NAME CGratton THarris RElliot MSmith MDavid for SCampbeli DATE 10/9/09 10/9/09 9 I 25 109 10/16/09 10/21/09 OFFICIAL RECORD COPY

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