Text

Basis for Proposed Changes. Part 1 of 2
	ML14335A624
Person / Time
Site:	Farley
Issue date:	11/24/2014
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML14335A689	List: ML14335A623; ML14335A629; NL-14-1385, Basis for Proposed Changes. Part 1 of 2; NL-14-1385, Joseph M. Farley Nuclear Plant - Units 1 and 2 - Request for Technical Specification Amendment Adoption of Previously NRC-Approved Generic Technical Specification Changes and Other Changes;
References
	NL-14-1385
	Download: ML14335A624 (156)
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Joseph M. Farley Nuclear Plant - Units 1 and 2 Request for Technical Specification Amendment Adoption of Previously NRC-Approved Generic Technical Specification Changes and Other Changes Enclosure 1 Basis for Proposed Changes

Enclosure 1 Basis for Proposed Changes Basis for Proposed Changes 1.0 Description The requested amendment will adopt various previously NRC-approved Technical Specifications Task Force (TSTF) Travelers and other changes to increase standardization. TSTF Travelers are generic changes chosen to increase the consistency between the Farley Technical Specifications, the Improved Standard Technical Specifications (ISTS) for Westinghouse plants (NUREG-1431), and the Technical Specifications of the other plants in the SNC fleet. The requested Travelers are:

1. TSTF-27-A, Revision 3, "Revise SR Frequency for Minimum Temperature for Criticality" (Page El-5)

2. TSTF-46-A, Revision 1, "Clarify the CIV Surveillance to Apply Only to Automatic Isolation Valves" (Page E1-9)

3. TSTF-87-A, Revision 2, "Revise 'RTBs Open' and 'CRDM De-energized' Actions to 'Incapable of Rod Withdrawal"' (Page E1-13)

4. TSTF-245-A, Revision 1, "AFW Train Operable When in Service" (Page E1-17)

5. TSTF-247-A, Revision 0, "Provide Separate Condition Entry for Each PORV and Block Valve" (Page E1-21)

6. TSTF-248-A, Revision 0, "Revise Shutdown Margin Definition for Stuck Rod Exception" (Page E1-25)

7. TSTF-266-A, Revision 3, "Eliminate the Remote Shutdown System Table of Instrumentation and Controls" (Page El -28)

8. TSTF-272-A, Revision 1, "Refueling Boron Concentration Clarification" (Page E1-31)

9. TSTF-273-A, Revision 2, "Safety Function Determination Program Clarifications" (Page E1-34)

10. TSTF-283-A, Revision 3, "Modify Section 3.8 Mode Restriction Notes" (Page E1-37)

11. TSTF-284-A, Revision 3, "Add 'Met vs. Perform' to Technical Specification 1.4, Frequency" (Page E1-43)

12. TSTF-308-A, Revision 1, "Determination of Cumulative and Projected Dose Contributions in RECP" (Page E1-47)

13. TSTF-312-A, Revision 1, "Administrative Control of Containment Penetrations" (Page E1-50)

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Enclosure 1 Basis for Proposed Changes

14. TSTF-314-A, Revision 0, "Require Static and Transient Fa Measurement" (Page E1-56)

15. TSTF-315-A, Revision 0, "Reduce Plant Trips Due to Spurious Signals to the NIS During Physics Testing" (Page E1-59)

16. TSTF-325-A, Revision 0, "ECCS Conditions and Required Actions with Less Than 100% Equivalent ECCS Flow" (Page El -63)

17. TSTF-340-A, Revision 3, "Allow 7-Day Completion Time for a Turbine AFW Pump Inoperable" (Page E1-66)

18. TSTF-343-A, Revision 1, "Containment Structural Integrity" (Page E1-69)

19. TSTF-349-A, Revision 1, "Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation" (Page E1-75)

20. TSTF-355-A, Revision 0, "Changes to RTS and ESF Tables" (Page E1-78)

21. TSTF-371-A, Revision 1, "NIS Power Range Channel Daily SR TS Change to Address Low Power Calibration" (Page E1-86)

22. TSTF-439-A, Revision 2, "Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet An LCO" (Page E1-90)

In addition, SNC is requesting to a change that reflects requirements in the Improved Standard Technical Specifications which were not added by a Traveler. These changes are:

23. ISTS Adoption #1 - Revise LCO 3.3.2 ESFAS Interlock P-4 Required Action Completion Time (Page E1-94)

SNC is requesting to adopt one change to increase the consistency between the Farley Technical Specifications and the other Westinghouse-designed plant in the SNC fleet, Vogtle Electric Generating Plant.

24. Revise LCO 3.5.5 to 8-hour Completion Time and Note allowance (Page E1-97) 2.0 Proposed Changes, Justifications, and No Significant Hazards Determinations Each Traveler is discussed in an individual analysis provided in Section 2.1 through 2.24. Each section contains the following topics:

Description of Proposed Change - This topic describes the effect of adopting the subject Traveler on the Farley Technical Specifications.

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Enclosure 1 Basis for Proposed Changes Differences Between the Proposed Change and the Approved Traveler -

This topic describes differences between the changes proposed to the Farley Technical Specifications and the ISTS mark-ups provided in the approved Traveler.

Summary of the Approved Traveler Justification - This topic summarizes the justification utilized by the NRC when approving the Traveler.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification - This topic describes any differences between the Traveler justification utilized by the NRC when approving the Traveler and the justification for adopting the Traveler in the Farley Technical Specifications.

License Commitments Required to Adopt this Change - Some Travelers require that licensees make regulatory commitments as a condition of adopting the change. This topic describes any such commitments being made by SNC as part of this request.

NRC Approval - This topic references the NRC letter, if any, approving the Traveler. It also provides example NRC approvals of plant-specific requests to adopt the Traveler. If the documents are in the NRC ADAMS system, the accession number (ACN) is given.

List of Affected Pages - This topic lists the Farley Technical Specification and Technical Specification Bases pages affected by the adoption of this Traveler.

Applicable Regulatory Requirements/Criteria - This topic describes how the justification satisfies the applicable regulatory requirements and criteria and provides a basis that the NRC staff may use to find the proposed amendment acceptable.

Significant Hazards Consideration - This topic provides an evaluation of whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment."

Each change not associated with a Traveler is discussed in an individual analysis provided in Sections 2.23 and 2.24. Each section contains the topics analogous to those in Sections 2.1 through 2.22, with the exception that the justification for the ISTS requirements or the NRC's approval of the Vogtle Technical Specification requirement is used as the basis.

The affected marked-up Technical Specifications pages are in Enclosure 2.

Retyped Technical Specification pages are in Enclosure 4.

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Enclosure 1 Basis for Proposed Changes Example mark-ups of the affected Technical Specification Bases pages are included for information only in Enclosure 3. The Bases will be revised under the Technical Specification Bases Control Program following NRC approval of the proposed Technical Specification changes.

To facilitate NRC review, each section will begin on a new page.

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Enclosure 1 Basis for Proposed Changes 2.1 TSTF-27-A, Revision 3, "Revise SR Frequency for Minimum Temperature for Criticality" Description of Proposed Change The proposed change revises the Frequency of Specification 3.4.2, "RCS Minimum Temperature for Criticality," SR 3.4.2.1, from "30 minutes thereafter," as modified by a Note which states, "Only required if low low Tavg alarm not reset and any RCS loop Tavg < 5470 F," to state, "In accordance with the Surveillance Frequency Control Program."

Differences Between the Proposed Change and the Approved Traveler The frequency for ISTS SR 3.4.2.1, and its associated Note, is modified by TSTF-27-A. The changes in TSTF-27-A would modify the Frequency for SR 3.4.2.1 to a periodic frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . As described in TS 5.5.21, Farley has adopted a Surveillance Frequency Control Program (SFCP) to control surveillances with periodic frequencies. The Frequency for SR 3.4.2.1, as modified by the changes identified in TSTF-27-A, will become a periodic frequency, and can be controlled under the SFCP. The Frequency for SR 3.4.2.1 is therefore modified to indicate that it is "In accordance with the Surveillance Frequency Control Program." The initial Frequency for this Surveillance will be 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The changes to SR 3.4.2.1 and the Bases for this SR are modified from that in TSTF-27-A to reflect this difference. NRC approval of the license change implementing the SFCP was provided in Amendment Numbers 185/180, dated July 18, 2011 (ACN ML11167A226).

Summary of the Approved Traveler Justification Specification 3.4.2, "RCS Minimum Temperature for Criticality," is designed to prevent criticality outside of the normal operating regime. There are no safety analyses that dictate the minimum temperature for criticality, but most low power accident analyses assume a specific starting temperature.

During the approach to criticality, reactor coolant system (RCS) temperature is closely watched. There are indications in the control room of deviations between actual and reference RCS temperature and on low RCS temperature to alert the operator if temperature is deviating from the program value. The Frequency of the SR only specifies how often temperature is logged, not how often it is watched. Therefore, the issue is whether or not the safety analysis assumptions are being protected, but how often RCS temperature is recorded in an operator's log. Therefore, this Traveler affects presentation and logging, not safety.

The current presentation can lead to inadvertently violating the SR Frequency with no effect on safety. The 30 minute SR Frequency "clock" continues even when RCS temperature is above the SR threshold or Applicability threshold temperature. Therefore, if temperature drops below the threshold value after more than 37 minutes (30 minutes + 25%) from the last time RCS temperature was logged, the SR Frequency has been violated. If temperature has unexpectedly decreased, the operator's attention should be on restoring temperature, not logging a value to meet a Surveillance. The operator is faced with making a decision of whether to focus his attention on the plant or on an administrative requirement. This is clearly adverse to safety. The other option is El - 5

Enclosure 1 Basis for Proposed Changes to perform the surveillance every 30 minutes until temperature is well above the threshold value in order to ensure that the SR has been performed if temperature should drop. This is not a beneficial use of an operator's time during the critical phases of a startup.

The proposed Frequency for SR 3.4.2.1 is modified to indicate that it is "In accordance with the Surveillance Frequency Control Program. The initial Frequency for this Surveillance will be 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This will ensure that Tavg iS logged at appropriate intervals (in addition to strip chart recorders and computer logging of temperature).

The requirement that RCS temperature must be above a certain value when the reactor is critical is stated in the LCO. This requirement will be monitored based on operating necessity whether or not it is specified in~a Surveillance Requirement. Requiring that the value be logged based on conditional circumstances is poor human-factors design and diverts the operator's attention from his duties without a compensating safety benefit.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not issue a letter approving TSTF-27-A, Revision 3; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. TSTF-27-A, Revision 3 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML021200265).

List of Affected Pages 3.4.2-1 B3.4.2-3 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 28, Reactivity Limits, states:

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, El -6

Enclosure 1 Basis for Proposed Changes changes in reactor coolant temperature and pressure, and cold water.

addition.

There is no regulatory requirement that specifies the interval between measurement of reactivity parameters, such as RCS temperature. The ISTS for Westinghouse Plants (NUREG-1431), Revision 3.1, provides for a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval. The proposed Surveillance Frequency is consistent with the NUREG-1431 Frequency.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the Surveillance Frequency for monitoring RCS temperature to ensure the minimum temperature for criticality is met.

The Frequency is changed from a 30 minute Frequency when certain conditions are met to a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program. The measurement of RCS temperature is not an initiator of any accident previously evaluated. The minimum RCS temperature for criticality is not changed. As a result, the mitigation of any accident previously evaluated is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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Enclosure 1 Basis for Proposed Changes

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the Surveillance Frequency for monitoring RCS temperature to ensure the minimum temperature for criticality is met.

The current, condition based Frequency represents a distraction to the control room operator during the critical period of plant startup. RCS temperature is closely monitored by the operator during the approach to criticality and temperature is recorded on charts and computer logs.

Allowing the operator to monitor temperature as needed by the situation and logging RCS temperature at a periodic Frequency that it is controlled in accordance with the Surveillance Frequency Control Program is sufficient to ensure that the LCO is met while eliminating a diversion of the operator's attention. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.2 TSTF-46-A, Revision 1, "Clarify the CIV Surveillance to Apply Only to Automatic Isolation Valves" Description of Proposed Change The proposed change modifies SR 3.6.3.4 and its associated Bases to delete the requirement to verify the isolation time of "each power operated" containment isolation valve and only require verification of each "automatic power operated isolation valve."

Differences Between the Proposed Change and the Approved Traveler The Farley Section 3.6.3 SR numbers are different from the ISTS Section 3.6.3 SR numbers. Farley SR 3.6.3.4 is equivalent to SR 3.6.3.5 in the ISTS. This has no effect on the requested change.

Summary of the Approved Traveler Justification ISTS SR 3.6.3.5 requires verification that the isolation time of "each power operated and each automatic containment isolation valve is within limits." The Bases for this SR state that the "isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis."

However, there are some valves credited as containment isolation valves that are power operated (i.e., can be remotely operated) that do not receive a containment isolation signal (e.g., a GDC 57 penetration). These power operated valves do not have an isolation time that is assumed in the accident analyses since they require operator action. The revised SR will clarify that it is only containment isolation valves (CIVs) that receive an automatic isolation signal that are in the scope of the SR. The associated Technical Specification Bases are also revised to reflect these changes. Deleting the reference to "power operated" isolation valve time testing reduces the potential for misinterpreting the requirements of this SR while maintaining the assumptions of the accident analysis.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None.

Licensee Commitments Required to Adopt this Change None.

NRC Approval The NRC did not issue a letter approving TSTF-46-A, Revision 1; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. TSTF-46-A, Revision 1 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML021200265). An example of a plant-specific NRC approval of the changes in TSTF-46-A is Peach Bottom Atomic Power Station, Units 2 and 3, Amendment Numbers 259/262 May 10, 2006 (ACN ML061070292).

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Enclosure 1 Basis for Proposed Changes List of Affected Pages 3.6.3-6 B3.6.3-12 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 16, Containment Design, states:

Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

In accordance with the requirement of Criterion 16 for an essentially leak-tight containment barrier, CIVs are designed to either close automatically when required or are capable of being closed manually if the valve required to be operated. It is not necessary to verify closure times for CIVs that do not receive an automatic isolation signal, and for which no closure time is assumed in the accident analysis.

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(C), states:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The change affects power operated CIVs that do not receive a containment isolation signal, and that do not have an isolation time that is assumed in the accident analyses, since they require operator action. There is no regulatory requirement to establish or verify isolation times for CIVs that are not credited to automatically close in the accident analysis. The changes will not alter the CIV design, or the design of the isolation logic or circuitry. The CIVs will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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Enclosure 1 Basis for Proposed Changes Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the requirements in Technical Specification SR 3.6.3.4, and the associated Bases, to delete the reference to verifying the isolation time of "each power operated" containment isolation valve (CIV) and only require verification of each "automatic power operated containment isolation valve." The closure times for CIVs that do not receive an automatic closure signal are not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The CIVs are used to respond to accidents previously evaluated. Power operated CIVs that do not receive an automatic closure signal are not assumed to close in a specified time. The proposed change does not change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not result in a change in the manner in which the CIVs provide plant protection or introduce any new or different operational conditions. Periodic verification that the closure times for CIVs that receive an automatic closure signal are within the limits established by the accident analysis will continue to be performed under SR 3.6.3.4. The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. There are also no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change provides clarification that only CIVs that receive an automatic isolation signal are within the scope of the SR 3.6.3.4. The proposed change does not result in a change in the manner in which the CIVs provide plant protection. Periodic verification that closure times for CIVs that receive an automatic isolation signal are within the limits established by the accident analysis will continue to be performed. The El - 11

Enclosure 1 Basis for Proposed Changes proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any Safety Analysis Limit.

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.3 TSTF-87-A, Revision 2, "Revise "RTBs Open" and "CRDM De-energized" Actions to "Incapable of Rod Withdrawal" Description of Proposed Change The proposed change modifies Specification 3.4.5, "RCS Loops - Mode 3,"

Required Actions C.2 and D. 1, from "De-energize all control rod drive mechanisms" to "Place the Rod Control System in a condition incapable of rod withdrawal." It also modifies Specification 3.4.9, "Pressurizer," Required Action A. 1, from requiring reactor trip breakers (RTBs) to be open after reaching MODE 3 to "Place the Rod Control System in a condition incapable of rod withdrawal,"

and to require full insertion of all rods.

Differences Between the Proposed Change and the Approved Traveler None.

Summary of the Approved Traveler Justification This change provides for a consistent presentation of the Required Actions. The specific method for ensuring that rods cannot be withdrawn is removed from the Technical Specifications. Since the revised Actions still assure rod withdrawal is precluded, this detail is not required to be in the TS to provide adequate protection of the public health and safety. There is no overall effect from the change. The requirement that the control rods are inserted and are not capable of being withdrawn is maintained. Therefore, removing this detail from the TS is acceptable.

This change (allowing alternate options to preclude rod withdrawal) is necessary to eliminate undesirable secondary effects of opening the RTBs. By opening the RTBs, plant interlock P-4 is tripped, which results in a trip of the main turbine and will close the main and bypass feedwater lines if RCS Tavg is below the low setpoint in MODE 3. Forcing reliance on AFW in this condition is not the intent, nor is it desirable, over continued use of normal feedwater. Additionally, Condition C of LCO 3.4.5 and LCO 3.9.1 are modified to reflect the LCO. The status of the reactor trip breakers is not a requirement of the LCO; and is therefore inappropriate in the Condition. No technical changes result from this change.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None.

Licensee Commitments Required to Adopt this Change None.

NRC Approval The NRC did not issue a letter approving TSTF-87-A, Revision 2; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. TSTF-87-A, Revision 2 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN 021200265).

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Enclosure 1 Basis for Proposed Changes List of Affected Pages 3.4.5-2 3.4.9-1 B3.4.5-1 B3.4.5-2 B3.4.5-4 B3.4.5-5 B3.4.9-3 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 28, Reactivity Limits, states:

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

There is no regulatory requirement that specifies the manner by which actions taken to prevent inadvertent rod withdrawal must be performed. As a result, it is not necessary to restrict the methods used to perform this function in order to meet Criterion 28.

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(B), states:

Criterion2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3.4.9, "Pressurizer," satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(C), states:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The intent of Required Actions in LCO 3.4.5 and LCO 3.4.9 directing, "De-energize all control rod drive mechanisms" is to prevent introduction of positive reactivity by inadvertent rod withdrawal. Changing this direction to state, "Place El - 14

Enclosure 1 Basis for Proposed Changes the Rod Control System in a condition incapable of rod withdrawal" satisfies the same intent in a less specific manner. There will be no changes to the design of the Reactor Coolant System or Rod Control System such that compliance with any of the regulatory requirements and guidance documents above would come into question. The Reactor Coolant System and Rod Control System will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

This change revises the Required Actions for LCO 3.4.5, "RCS Loops -

Mode 3," Conditions C.2 and D.1, from "De-energize all control rod drive mechanisms," to "Place the Rod Control System in a condition incapable of rod withdrawal." It also revises LCO 3.4.9, "Pressurizer," Required Action A. 1, from requiring the Reactor Trip Breakers to be open after reaching MODE 3 to "Place the Rod Control System in a condition incapable of rod withdrawal," and to require full insertion of all rods.

Inadvertent rod withdrawal can be an initiator for design basis accidents or events during certain plant conditions, and therefore must be prevented under those conditions. The proposed Required Actions for LCO 3.4.5 and LCO 3.4.9 satisfy the same intent as the current Required Actions, which is to prevent inadvertent rod withdrawal when an applicable Condition is not met, and is consistent with the assumptions of the accident analysis. As a result, the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not change how the plant would mitigate an accident previously evaluated as in both the current and proposed requirements, rod withdrawal is prohibited. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

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Enclosure I Basis for Proposed Changes The proposed change provides less specific, but equivalent, direction on the manner in which inadvertent control rod withdrawal is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9 are not met. Rod withdrawal will continue to be prevented when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change provides the operational flexibility of allowing alternate, but equivalent, methods of preventing rod withdrawal when LCO 3.4.5 and LCO 3.4.9 are not met. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.4 TSTF-245-A, Revision 1, "AFW Train Operable When in Service" Description of Proposed Change The proposed TS modifies Surveillance Requirement 3.7.5.1, 3.7.5.3, and 3.7.5.4 to add a Note stating that "AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation."

Differences Between the Proposed Chan-ge and the Approved Traveler Farley SR 3.7.5.1 contains a Note stating that the SR is, "Not required to be performed for the AFW flow control valves when < 10% RTP or when the AFW flow control system is not in manual control." This Note does not appear in the ISTS, and functionally serves the same purpose as the Note that is added in TSTF-245-A. The existing Note in SR 3.7.5.1 is deleted and replaced with the SR Note from TSTF-245-A, and conforming changes to the Bases text are made to reflect deletion of the plant-specific Note.

ISTS SR 3.7.5.3 contains a note stating that the SR is "Not applicable in MODE 4 when steam generator is relied upon for heat removal." The approved Traveler deletes and replaces this note. Farley SR 3.7.5.3 does not currently include this note, and will add the note identified in the approved Traveler under this change.

Summary of the Approved Traveler Justification Auxiliary Feedwater (AFW) is a dual use system. As such, AFW valves may be positioned other than that required for decay and residual heat removal during Modes I (below 10% Rated Thermal Power), 2, 3, 4, and 5, when the AFW system is being used to maintain steam generator level. Adding a Note stating that an AFW train may be considered operable during alignment and operation for steam generator level control, if capable of being manually realigned to the AFW mode of operation would clarify the intended dual-use flexibility of the AFW system and prevent unnecessary Action entry.

The Note provides an exception that allows the AFW system to be out of its normal standby alignment and temporarily incapable of automatic initiation without declaring the affected AFW train(s) inoperable. Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an expected function of the AFW system, operability (i.e., the intended safety function) should be maintained during these operations. Additionally, following a reactor trip, AFW flow provides the source of makeup to the steam generators. If excessive RCS cooldown is experienced and it is caused by a large amount of AFW flow, the Turbine Driven AFW Pump may be stopped in order to limit RCS cooldown.

However, the Turbine Driven AFW Pump would still remain available for steam generator level control and could be restored by the operator should the need arise.

NUREG-1431 incorporates the changes identified in TSTF-245-A, and includes a Note in SR 3.7.5.1, 3.7.5.3 and 3.7.5.4 stating that "AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level El - 17

Enclosure 1 Basis for Proposed Changes control, if it is capable of being manually realigned to the AFW mode of operation."

Farley Operating Procedures and Emergency Operator Procedures contain steps to support realignment of the AFW system from manual steam generator level control mode to the emergency operation mode if required.

With regard to the AFW system, the NRC staff has previously issued a determination' on the effects of manual operation on Operability of the AFW system, which concluded that manual operation does not render the AFW system inoperable, provided manual action can perform the same function. The NRC recognizes that AFW is a dual-use system and may be used during startup of the plant, normal shutdown, and hot standby conditions, and that it is control band operated during these conditions in the manual mode of operation. In such situations, the AFW system is considered Operable. The NRC letter can be found as an attachment to TSTF-245-A.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None.

Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not issue a letter approving TSTF-245-A, Revision 1; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. TSTF-245-A, Revision 1 has been adopted by many plants as part of complete conversion to the ISTS, such as Beaver Valley Power Station (ACN ML050610351). An example of a plant-specific NRC approval of the changes in TSTF-245-A is Comanche Peak Units 1 and 2, Amendment Numbers 126/126 dated April 4, 2006 (ACN ML060860258).

List of Affected Pages 3.7.5-2 3.7.5-3 B3.7.5-8 B3.7.5-9 Applicable Regqulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(B), states:

Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 Harold, Jeffrey F. (NRC) to Stephen E. King (Consolidated Edison), "Manual vs. Automatic Operation as it Relates to Auxiliary Feedwater Operability at Indian Point Nuclear Generating Unit No. 2 (TAC No. M98056), dated May 23, 1997.

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Enclosure 1 Basis for Proposed Changes The proposed TS changes would allow the AFW train(s) to be considered operable during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. In practice, this would allow the AFW valves may be positioned other than that required for decay and residual heat removal during Modes I (below 10% Rated Thermal Power), 2, 3, 4, and 5, when the AFW system is being used to maintain steam generator level. The decay and residual heat loads will be low during the low power and shutdown conditions where the Note would apply, and there is sufficient time to realign the AFW system from manual steam generator level control mode to the AFW mode if needed.

There will be no changes to the auxiliary feedwater system design such that compliance with the regulatory requirements and guidance document above would come into question. The auxiliary feedwater system will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the requirements in Technical Specification 3.7.5, "Auxiliary Feedwater (AFW) System," to clarify the operability of an AFW train when it is aligned for manual steam generator level control. The AFW System is not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The AFW System is used to respond to accidents previously evaluated. The proposed change does not affect the design of the AFW System, and no physical changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

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Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed).

The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Manual control of AFW level control valves is not an accident initiator. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change provides the operational flexibility of allowing an AFW train(s) to be considered operable when it is not in the normal standby alignment and is temporarily incapable of automatic initiation, such as during alignment and operation for manual steam generator level control, provided it is capable of being manually realigned to the AFW heat removal mode of operation. The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to supply water to the steam generators to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any Safety Analysis Limit.

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.5 TSTF-247-A, Revision 0, "Provide Separate Condition Entry for Each PORV and Block Valve" Description of Proposed Change The proposed change modifies Specification 3.4.11, "Pressurizer PORVs," to provide separate Condition entry for each PORV and each block valve.

Differences Between the Proposed Change and the Approved Traveler None. However, TSTF-247-A provides options depending on the number of PORV and block valves that are included in the plant design. The design of Farley, Units 1 and 2, includes two PORVs and associated block valves. The options from TSTF-247-A for plants with three PORVs and associated block valves are not adopted.

Summary of the Approved Traveler Justification The existing LCO 3.4.11 Conditions allow separate condition entry for each pressurizer power operated relief valve (PORV). The Conditions and Required Actions provide appropriate compensatory measures for separate condition entry for each inoperable PORV. The Conditions and Required Actions also provide appropriate compensatory actions for separate condition entry for each block valve. Therefore, the Actions Note is modified to allow separate condition entry for each block valve.

Condition F is modified to apply when both block valves are inoperable. The existing Actions are modified to no longer require that both PORVs be placed in manual control if both block valves are inoperable. This avoids a potential situation where a plant shutdown is required if one of the block valves cannot be restored within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the PORVs, which will be needed for Low Temperature overpressure protection, cannot perform their automatic pressure relief function. Deletion of Action F. 1 removes an unnecessary requirement since separate condition entry for each block valve makes it redundant with Action C. 1. Action F.3 is also eliminated (Restore remaining block valve(s) to operable status). With separate condition entry for each block valve this ACTION is no longer necessary.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification As described in TSTF-247-A, Condition F is modified to apply when both PORV block valves are inoperable, and the existing Required Actions are modified to not require that the PORVs be placed in manual control under these circumstances. The basis for this is that if the block valves are not restored within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months a plant shutdown is required, and the PORVs will be needed for low temperature overpressure protection (LTOP). Therefore, the PORVs should not be placed in manual control.

The PORVs are not currently credited as part of the Farley LTOP strategy.

TSTF-247-A is being implemented to preserve the PORVs as a potential defense-in-depth LTOP option for future use. Adopting these changes is appropriate because, as stated in the justification text for TSTF-247-A, "The Conditions and Required Actions also provide appropriate compensatory actions El - 21

Enclosure 1 Basis for Proposed Changes for separate condition entry for each block valve. Therefore, the Actions Note is modified to allow separate condition entry for each block valve."

Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not issue a letter approving TSTF-247-A, Revision 0; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. TSTF-247-A, Revision 2 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN 021200265). An example of a plant-specific NRC approval of the changes in TSTF-247-A is Callaway Unit 1, Amendment Number 188, dated November 25, 2008 (ACN ML082910895).

List of Affected Pages 3.4.11-1 3.4.11-3 B3.4.11-4 B3.4.11-6 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(C), states:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

There will be no changes to the pressurizer PORV or block valve design or operation such that compliance with any of the regulatory requirements and guidance documents above would come into question. There will be no changes to the plant design or operations such that compliance with any of the regulatory requirements and guidance documents above would come into question. The plant and its systems will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Siqnification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

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Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change revises the requirements in Technical Specification 3.4.11, "Pressurizer PORVs," to clarify that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the overpressure protection capabilities of the PORVs. The pressurizer block valves are used to isolate their respective PORV in the event it is experiencing excessive leakage, and are not an initiator of any design basis accident or event. Therefore the proposed change does not increase the probability of any accident previously evaluated. The PORV and block valves are used to respond to accidents previously evaluated. The proposed change does not affect the design of the PORV and block valves, and no physical changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Operation of the PORV block valves is not an accident initiator. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes provide clarification that separate Condition entry is allowed for each block valve. Additionally, the Actions are modified to no longer require that the PORVs be placed in manual operation when both block valves are inoperable and cannot be restored to operable status within the specified Completion Time. This preserves the El - 23

Enclosure 1 Basis for Proposed Changes overpressure protection capabilities of the PORVs. The proposed change does not result in a change in the manner in which the PORV and block valves provide plant protection. The PORVs will continue to provide overpressure protection, and the block valves will continue to provide isolation capability in the event a PORV is experiencing excessive leakage. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions.

The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.6 TSTF-248-A, Revision 0, "Revise Shutdown Margin Definition for Stuck Rod Exception" Description of Proposed Change The proposed change revises the definition of Shutdown Margin (SDM) to eliminate the requirement that shutdown margin calculations must assume the single rod cluster control assembly (RCCA) of highest worth is fully withdrawn if all RCCAs can be verified to be fully inserted by two independent means.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification The Shutdown Margin definition states, "SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures." The proposed change modifies the definition to include, "However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation."

The consideration of a stuck rod is provided in the SDM definition to allow for a single failure of one rod to not insert when a scram is initiated. However, with positive indication that all rods are already fully inserted, such a provision is overly conservative. This change is consistent with the definition of Shutdown Margin provided in NUREG-1431.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-248-A, Revision 0, in a letter from William D. Beckner (NRC) to Anthony R. Pietrangelo (NEI), dated October 31, 2000 (ACN ML003775261). An example of a plant-specific NRC approval of the changes in TSTF-248-A is Catawba Units 1 and 2, McGuire Units 1 and 2, and Oconee Units 1, 2, and 3 amendments, dated May 28, 2010 (ACN ML101390415).

List of Affected Pages 1.1-5 El - 25

Enclosure 1 Basis for Proposed Changes Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criteria:

Criterion 25, Protection System Requirements for Reactivity Control Malfunctions, states:

The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

Criterion 27, Combined Reactivity Control Systems Capability, states:

The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

Typically the shutdown margin calculations assume the most reactive control rod fails insert into the core (i.e., a stuck rod). However, when it can be confirmed by two independent methods that all rods are inserted, it is not appropriate to include a margin for stuck rods.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The method for calculating shutdown margin is not an initiator of any accident previously evaluated. If it can be verified by two independent means that all control rods are inserted, the calculated Shutdown Margin, without the conservatism of assuming the highest worth control rod is withdrawn, is accurate and consistent with the assumptions in the accident analysis. As El - 26

Enclosure 1 Basis for Proposed Changes a result, the mitigation of any accident previously evaluated is not affected. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change modifies the definition of Shutdown Margin to eliminate the requirement to assume the highest worth control rod is fully withdrawn when calculating Shutdown Margin if it can be verified by two independent means that all control rods are inserted. The additional margin of safety provided by the assumption that the highest worth control rod is fully withdrawn is unnecessary if it can be independently verified that all controls rods are inserted. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.7 TSTF-266-A, Revision 3, "Eliminate the Remote Shutdown System Table of Instrumentation and Controls" Description of Proposed Change The proposed change relocates the list of Remote Shutdown System instrumentation and controls in Specification 3.3.4, "Remote Shutdown System,"

from the Technical Specifications to the Technical Specification Bases.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification This change eliminates the table of instrumentation and controls referenced in the Specification for the Remote Shutdown System. The specific instruments and controls necessary for each Function provided by the Remote Shutdown System are currently listed in a Table in the Specifications. This change will eliminate the table and the information will be placed in the Bases. It is unnecessary to list the specific instruments and controls in the Technical Specifications to provide adequate assurance that the functions can be performed. GDC 19 requires that the remote shutdown capability be provided. The LCO provides references to the Functions, which are described in the Bases. This is sufficient to ensure that the system will be operable. Listing the specific instrumentation and controls is unnecessary and may lead to needless expenditure of licensee and NRC resources processing license amendments to revise the table when the information can be adequately controlled by the licensee.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-266-A, Revision 3, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated September 10, 1999 (ACN 9909160189). An example of a plant-specific NRC approval of the changes in TSTF-266-A is South Texas Project, Units 1 and 2, Amendment Numbers 163/152 dated August 20, 2004 (ACN ML042370841).

List of Affected Pages 3.3.4-1 3.3.4-3 B 3.3.4-2 B 3.3.4-3 B 3.3.4-6 El - 28

Enclosure 1 Basis for Proposed Changes Applicable Regqulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 19, Control Room, states:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Criterion 19 requires that remote shutdown capability be provided. The Remote Shutdown System Functions are described in the Updated Final Safety Analysis Report. The definition of operable in the Farley specifications states that a system shall be operable or have operability when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system to perform its specified safety function(s) are also capable of performing their related support function. This definition provides adequate guidance for determining what instrumentation and controls are necessary for a particular remote shutdown function. The ability to transfer control of a function from the control room to the remote shutdown panel is a required support function by the definition of operable. Therefore, listing specific instrumentation and controls is unnecessary and may lead to needless expenditure of licensee and NRC resources processing license amendments to revise the Remote Shutdown System details in the Technical Specifications when these details are not necessary to describe the actual regulatory requirements.

Therefore, they can be removed from the Technical Specifications and placed in the Bases without a significant impact on safety.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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Enclosure 1 Basis for Proposed Changes

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places them in the Bases. The Technical Specifications continue to require that the instrumentation and controls be operable. The location of the list of Remote Shutdown System instrumentation and controls is not an initiator to any accident previously evaluated. The proposed change will have no effect on the mitigation of any accident previously evaluated because the instrumentation and controls continue to be required to be operable.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change removes the list of Remote Shutdown System instrumentation and controls from the Technical Specifications and places it in the Bases. The review performed by the NRC when the list of Remote Shutdown System instrumentation and controls is revised will no longer be needed unless the criteria in 10 CFR 50.59 are not met such that prior NRC review is required. The Technical Specification requirement that the Remote Shutdown System be operable, the definition of operability, the requirements of 10 CFR 50.59, and the Technical Specifications Bases Control Program are sufficient to ensure that revision of the list without prior NRC review and approval does not introduce a significant safety risk. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.8 TSTF-272-A, Revision 1, "Refueling Boron Concentration Clarification" Description of Proposed Change The proposed change adds an Applicability Note to Specification 3.9.1, "Boron Concentration." The Note clarifies that the LCO only applies to the refueling canal and the refueling cavity when those volumes are connected to the Reactor Coolant System.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification Specification 3.9.1, "Boron Concentration," is revised to clarify that the boron concentration limits do not apply to the refueling canal and refueling cavity when these areas are not connected to the RCS. This Specification limits the boron concentrations of the RCS, the refueling canal, and the refueling cavity during refueling to ensure that the reactor remains subcritical. However, when the refueling canal and refueling cavity are isolated from the RCS, no potential for dilution exists. In this condition it is not necessary to place a limit on the boron concentration of water in the refueling cavity and the refueling canal. This change is consistent with the intent of the Specification and eliminates restrictions that have no effect on safety.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-272-A, Revision 1, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated December 21, 1999 (ACN ML993630256). An example of a plant-specific NRC approval of the changes in TSTF-272-A is Millstone Unit 2, Amendment Number 263, dated January 11, 2002 (ACN ML013440338).

List of Affected Pages 3.9.1-1 B3.9.1-3 B3.9.1-4 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criteria:

Criterion 28, Reactivity Limits, states:

The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of El - 31

Enclosure 1 Basis for Proposed Changes postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

Criterion 61, Fuel Storage and Handling and Radioactivity Control, states:

The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

The proposed change clarifies that the limits on Reactor Coolant System boron concentration are only applicable to those portions of the Reactor Coolant System that are in communication with the reactor core and can, therefore, affect core reactivity.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change modifies the Applicability of Specification 3.9.1, "Boron Concentration," to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the Reactor Coolant System (RCS). The boron concentration of water volumes not connected to the RCS are not an initiator of an accident previously evaluated. The ability to mitigate any accident previously evaluated is not affected by the boron concentration of El - 32

Enclosure 1 Basis for Proposed Changes water volumes not connected to the RCS. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change modifies the Applicability of Specification 3.9.1, "Boron Concentration," to clarify that the boron concentration limits are only applicable to the refueling canal and the refueling cavity when those volumes are attached to the RCS. Technical Specification SR 3.0.4 requires that Surveillances be met prior to entering the Applicability of a Specification. As a result, the boron concentration of the refueling cavity or the refueling canal must be verified to satisfy the LCO prior to connecting those volumes to the RCS. The margin of safety provided by the refueling boron concentration is not affected by this change as the RCS boron concentration will continue to satisfy the LCO. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.9 TSTF-273-A, Revision 2, "Safety Function Determination Program Clarifications" Description of Proposed Chanae The proposed TS changes adds explanatory text to the LCO 3.0.6 Bases clarifying the "appropriate LCO for loss of function," and that consideration does not have to be made for a loss of power in determining loss of function.

Explanatory text is also added to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.15 to provide clarification of these same issues.

Differences Between the Proposed Change and the Approved Traveler None.

Summary of the Approved Traveler Justification TS 5.5.15, "Safety Function Determination Program," implements the requirements of LCO 3.0.6. The SFDP program description in TS 5.5.15 is revised to clarify in the requirements that consideration does not have to be made for a loss of power in determining loss of function. TS 5.5.15 is also revised to incorporate an editorial change for consistency in meaning. The Bases for LCO 3.0.6 is revised to provide clarification of the "appropriate LCO for loss of function," and that consideration does not have to be made for a loss of power in determining loss of function.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval TSTF-273-A, Revision 2, was approved by the NRC as documented in a letter from William Beckner (NRC) to James Davis (NEI), dated August 16, 1999.

TSTF-273-A, Revision 2 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML021200265).

An example of a plant-specific NRC approval of the changes in TSTF-273-A, Revision 2 is Susquehanna Steam Electric Station, Units 1 and 2, Amendment Numbers 209/183 dated February 25, 2003 (ACN ML060860258).

List of Affected Pages 5.5-13 B3.0-10 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow El - 34

Enclosure 1 Basis for Proposed Changes any remedial action permitted by the technical specifications until the condition can be met.

The SFDP, as described in TS 5.5.15, implements the requirements of Limiting Condition for Operation (LCO) 3.0.6, and ensures that loss of safety function is detected and appropriate actions are taken. There will be no changes to the plant design or operation such that compliance with the regulatory requirements and guidance document above would come into question. The plant and its systems will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed TS changes add explanatory text to the programmatic description of the Safety Function Determination Program (SFDP) in Specification 5.5.15 to clarify in the requirements that consideration does not have to be made for a loss of power in determining loss of function.

The Bases for LCO 3.0.6 is revised to provide clarification of the "appropriate LCO for loss of function," and that consideration does not have to be made for a loss of power in determining loss of function. The changes are editorial and administrative in nature, and therefore do not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes are editorial and administrative in nature and do not result in a change in the manner in which the plant operates. The loss of function of any specific component will continue to be addressed in its El - 35

Enclosure 1 Basis for Proposed Changes specific TS LCO and plant configuration will be governed by the required actions of those LCOs. The proposed changes are clarifications that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to TS 5.5.15 are clarifications and are editorial and administrative in nature. No changes are made the LCOs for plant equipment, the time required for the TS Required Actions to be completed, or the out of service time for the components involved. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed changes will not result in plant operation in a configuration outside the design basis. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.10 TSTF-283-A, Revision 3, "Modify Section 3.8 Mode Restriction Notes" Description of Proposed Change The proposed change revises several Specification 3.8.1, "AC Sources -

Operating," Surveillance Notes to allow full or partial performance of the SRs to reestablish Operability provided an assessment determines the safety of the plant is maintained or enhanced. These Surveillances currently have Notes prohibiting their performance in Modes 1 or 2, or in Modes 1, 2, 3, or 4.

SR 3.8.1.7, which tests the transfer of Alternating Current (AC) sources from normal to alternate offsite circuits, contains a Note prohibiting performance in Mode 1 or 2. The Note is modified to state that performance is normally prohibited in Mode 1 or 2, but may be performed to reestablish Operability provided an assessment determines the safety of the plant is maintained or enhanced.

SR 3.8.1.9, which tests the response to a loss of offsite power signal, contains a Note prohibiting performance in Mode 1, 2, 3, or 4. The Note is modified to state that performance is normally prohibited in Mode 1, 2, 3, or 4, but portions of the SR may be performed to reestablish Operability provided an assessment determines the safety of the plant is maintained or enhanced.

SR 3.8.1.14, which verifies the transfer from Diesel Generator (DG) to offsite power, contains a Note prohibiting performance in Mode 1, 2, 3, or 4. The Note is modified to state that performance is normally prohibited in Mode 1, 2, 3, or 4, but portions of the SR may be performed to reestablish Operability provided an assessment determines the safety of the plant is maintained or enhanced.

SR 3.8.1.17, which verifies the response to a loss of offsite power signal and Engineered Safety Feature (ESF) actuation signal, contains a Note prohibiting performance in Mode 1, 2, 3, or 4. The Note is modified to state that performance is normally prohibited in Mode 1, 2, 3, or 4, but portions of the SR may be performed to reestablish Operability provided an assessment determines the safety of the plant is maintained or enhanced.

Differences Between the Proposed Change and the Approved Traveler Several of the changes approved in TSTF-283-A are not applicable to the Farley Technical Specifications. As identified in the following table, Farley SRs 3.8.1.8, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.15, 3.8.1.16, and 3.8.1.18 do not include a Note restricting the MODES in which the SR may be performed. As such, addition of the TSTF Note that provides an exception to the MODE restrictions is not necessary, and is not adopted. The MODE restriction Notes for these SRs, which appear in NUREG-1431, Rev. 1, were not adopted during the Farley ITS conversion because they were not part of the plant licensing basis that was described in the plant Custom Technical Specifications.

Additionally, several previous plant-specific amendments to adopt TSTF-283-A withdrew the proposed changes to the Specification 3.8.4 Surveillances in response to NRC questions. The following table summarizes the differences between the TSTF-283-A proposed Surveillance Note changes to the El - 37

Enclosure 1 Basis for Proposed Changes Westinghouse plant ISTS and the proposed Surveillance Note changes to the Farley Technical Specifications:

TSTF-283-A Affected Surveillance Equivalent Farley Surveillance and and Description Proposed Change SR 3.8.1.8 (transfer of AC sources from SR 3.8.1.7 - Adopt as proposed in normal to alternate offsite circuit) TSTF-283-A.

SR 3.8.1.9 (largest post-accident load SR 3.8.1.8 - Not applicable. Current reject) SR does not have Mode restriction Note.

SR 3.8.1.10 (load reject) SR 3.8.1.18 - Not applicable. Current SR does not have Mode restriction Note.

SR 3.8.1.11 (response to loss of offsite SR 3.8.1.9 - Adopt as proposed in power signal) TSTF-283-A.

SR 3.8.1.12 (response to an ESF SR 3.8.1.10 - Not applicable. The actuation signal) existing Farley SR specifies which portions of the SR are restricted in Modes 1 and 2.

SR 3.8.1.13 (verify DG automatic trips SR 3.8.1.11 - Not applicable. Current are bypassed) SR does not have Mode restriction Note.

SR 3.8.1.14 (DG 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months run) SR 3.8.1.12 - Not applicable. Current SR does not have Mode restriction Note.

SR 3.8.1.16 (verify transfer from DG to SR 3.8.1.14 - Adopt as proposed in offsite power) TSTF-283-A.

SR 3.8.1.17 (verify DG transfer from SR 3.8.1.15 - Not applicable. Current test mode) SR does not have Mode restriction Note.

SR 3.8.1.18 (verify interval between SR 3.8.1.16 - Not applicable. Current sequencer load blocks) SR does not have Mode restriction Note.

SR 3.8.1.19 (response to loss of offsite SR 3.8.1.17 - Adopt as proposed in power signal and ESF actuation signal) TSTF-283-A.

SR 3.8.4.6 (battery charger capacity) SR 3.8.4.6 - Not applicable. Current SR has a similar allowance.

SR 3.8.4.7 (battery capacity) SR 3.8.4.7 - Not requested.

SR 3.8.4.8 (battery discharge test) SR 3.8.4.8 - Not requested.

The TSTF-283-A Bases changes incorrectly state that the associated Notes restrict performance of the Surveillances in Mode 1 and 2. Several of the Surveillances restrict performance of the Surveillances in Mode 1, 2, 3, or 4. This error is corrected in the Farley Bases.

Summary of the Approved Traveler Justification The proposed changes to Specification 3.8.1 will potentially avoid a plant shutdown if corrective maintenance (planned or unplanned) performed during power operation results in the need to perform any of the revised Surveillances to El - 38

Enclosure 1 Basis for Proposed Changes demonstrate Operability. The proposed changes do not affect either the frequency of conducting the SRs, the surveillance to be performed, or the performance criteria specified in the SRs. The only change is to the reactor modes during which the surveillance may be performed.

The allowance to perform the Surveillances in currently prohibited Modes is restricted to only allow the Surveillances to be performed for the purpose of reestablishing Operability (e.g. post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated Operability concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed.

Note that the Maintenance Rule provision contained in 10 CFR 50.65(a)(4) states that before performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. This includes the performance of Surveillances to reestablish Operability. Therefore, in addition to the assessment required by the Surveillance Notes, an assessment of plant risk will also be performed.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not document their approval of TSTF-283-A, Revision 3 in a letter but it was incorporated into Revision 2 of the ISTS NUREGs. An example of plant-specific NRC approval of the changes in TSTF-283-A are Palo Verde Units 1, 2, and 3, Amendment Numbers 156/156/156, dated September 29, 2005 (ACN ML052720567); Diablo Canyon Units 1 and 2, Amendment Numbers 174/176, dated September 28, 2004 (ACN ML042790255); and Columbia Generating Station, Amendment Number 204 dated May 1, 2007 (ACN ML070920469). In these example plant-specific NRC approvals, the requested changes to the Farley Technical Specifications were approved by the NRC.

List of Affected Pages 3.8.1-8 3.8.1-9 3.8.1-12 3.8.1-13 B3.8.1-20 El - 39

Enclosure 1 Basis for Proposed Changes B3.8.1-22 B3.8.1-25 B3.8.1-27 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criteria:

Criterion 17, Electric Power Systems, states:

An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.

This criterion requires provisions to minimize the probability of losing electric power from the remaining electric power supplies as a result of loss of power from the unit, the offsite transmission network, or the onsite power supplies.

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Enclosure 1 Basis for Proposed Changes Criterion 18, Inspection and Testing of Electric Power Systems, states:

Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.

10 CFR 50.36(c)(3), "Technical Specifications," requires a licensee's Technical Specifications to have Surveillance Requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operations are within safety limits, and that the limiting conditions for operation (LCOs) will be met. The Surveillance Requirements may include mode restrictions based on the safety aspects of conducting the surveillances in excluded modes.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change modifies Mode restriction Notes on four diesel generator (DG) Surveillances to allow performance of the Surveillance in whole or in part to reestablish DG Operability. The emergency diesel generators and their associated emergency loads are accident mitigating features, and are not an initiator of any accident previously evaluated. As a result the probability of any accident previously evaluated is not increased. The proposed change allows Surveillance testing to be performed in whole or in part to reestablish Operability of a DG. The consequences of an accident previously evaluated during the period that the DG is being tested to reestablish Operability are no different from the consequences of an accident previously evaluated while the DG is El - 41

Enclosure 1 Basis for Proposed Changes inoperable. As a result, the consequences of any accident previously evaluated are not increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The purpose of Surveillances is to verify that equipment is capable of performing it's assumed safety function. The proposed change will only allow the performance of the Surveillances to reestablish Operability and the proposed changes may not be used to remove a DG from service. In addition, the proposed change will potentially shorten the time that a DG is unavailable because testing to reestablish Operability can be performed without a plant shutdown. The proposed changes also require an assessment to verify that plant safety will be maintained or enhanced by performance of the Surveillance in the normally prohibited Modes.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.11 TSTF-284-A, Revision 3, "Add 'Met vs. Perform' to Technical Specification 1.4, Frequency" Description of Proposed Change The change inserts a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that utilize the terms "met" and "perform." The changes also modify SR 3.4.11.1, SR 3.4.12.5, and SR 3.9.3.2 to appropriately use "met" and "perform" exceptions.

Differences Between the Proposed Change and the Approved Traveler TSTF-284-A, Revision 3 includes changes to SR 3.1.11.1 and SR 3.1.11.2 of ISTS Specification 3.1.11, "SDM Test Exceptions." This LCO allows suspension of SDM requirements in MODE 2 provided specific conditions are met in order facilitate measurement control rod worth and SOM. The Farley Technical Specifications do not include a Specification that is analogous to ISTS TS 3.1.11, "SDM Test Exceptions," or SRs that are analogous to ISTS SRs 3.1.11.1 and 3.1.11.2. Therefore, the TS and Bases changes identified in TSTF-284-A for ISTS 3.1.11 are not adopted.

The Farley Section 3.9 specification numbers are different from the ISTS Section 3.9 specification numbers. Farley Specification 3.9.3, "Containment Penetrations," is equivalent to Specification 3.9.4 in the ISTS. This has no effect on the requested change.

Changes to the Actions Bases for Specification 3.4.11, "Pressurizer PORVs," are not adopted. The changes described in the TSTF are related to a Note in the ISTS that provides an exception to LCO 3.0.4 that allows entry into MODES 1, 2, and 3 to perform cycling of the PORVs or block valves in order to demonstrate their operability. Consistent with NUREG-1431, Farley Technical Specification 3.4.11, and its associated Bases, do not include the Note providing this exception to LCO 3.0.4.

Changes identified in TSTF-284-A to ISTS SR 3.4.11.2, and its associated Bases, are also not adopted. Farley SR 3.4.11.2 requires performance of a complete cycle of each PORV during MODE 3 or 4, and is modified by a Note stating that this testing is not required to be performed prior to entering MODE 3.

This is substantially different from the requirements in analogous ISTS SR 3.4.11.2, which requires performance of a complete cycle of each PORV, but does not specify the MODES in which it must be performed. These differences are such that the changes in TSTF-284-A are not appropriate.

TSTF-284-A identifies changes to ISTS 3.4.12, "LTOP System," SR 3.4.12.8.

This SR requires performance of a Channel Operational Test (COT) for each required PORV, excluding actuation. The Farley Technical Specifications do not include an SR that is analogous to ISTS SR 3.4.12.8. Therefore, the TS and Bases changes identified in TSTF-284-A for ISTS SR 3.4.12.8 are not adopted.

Summary of the Approved Traveler Justification The change inserts a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that El -43

Enclosure 1 Basis for Proposed Changes utilize the terms "met" and "perform." The changes also modify SRs as necessary to appropriately use "met" and "perform" exceptions. The added examples parallel the existing example 1.4-3 of Notes that allow for the SR to be "Not required to be performed. . .". The examples will alleviate misunderstanding and provide explicit direction for these types of SR Notes.

NUREG-1433 (BWR/4 plants) and -1434 (BWR/6 plants) contain a discussion in Specification 1.4 regarding the use of "met" and "performed" in SR Notes.

Similarly, the Writers Guide provides a distinction between these phrases.

However, NUREG-1430 (B&W), -1431 (Westinghouse), and -1432 (Combustion Engineering) did not originally contain this detail, even though various locations throughout these NUREGs provide Notes with "met" and "performed" distinctions.

Inserting this material will provide for better use, application, and understanding of these Notes. Furthermore, this change will establish consistency between the NUREGs. With this clarification, several exceptions that are unclear or have incorrect usage of "met" and "perform" are also corrected.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval TSTF-284-A, Revision 3, was approved by the NRC as documented in a letter from William Beckner (NRC) to James Davis (NEI), dated February 16, 2000 (ACN ML003684596). TSTF-284-A, Revision 3 has been adopted by many plants as part of complete conversion to the ISTS, such as Beaver Valley Power Station (ACN ML070160593). An example of a plant-specific NRC approval of the changes in TSTF-284-A, Revision 3 is Columbia Generating Station, Amendment Number 205 dated December 13, 2007 (ACN ML073120270).

List of Affected Pages 1.4-1 1.4-4 3.4.11-3 3.4.12-4 3.9.3-2 B3.4.11-7 B3.4.11-8 B3.4.12-11 B3.9.3-6 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow El -44

Enclosure 1 Basis for Proposed Changes any remedial action permitted by the technical specifications until the condition can be met.

The changes insert a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that utilize "met" and "perform." With this clarification, several exceptions that are unclear or have incorrect usage of "met" and "perform" are also corrected. The changes also modify SR 3.4.11.1, SR 3.4.12.5, and SR 3.9.3.2 to appropriately use "met" and "perform" exceptions. The changes to LCO 1.4 clarify implementation of the requirements for LCOs that have "met" or "performed" exceptions. There will be no changes to the plant design or operations such that compliance with any of the regulatory requirements and guidance documents above would come into question. The plant and its systems will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes insert a discussion paragraph into Specification 1.4, and several new examples are added to facilitate the use and application of SR Notes that utilize the terms "met" and "perform. The changes also modify SRs in multiple Specifications to appropriately use "met" and "perform" exceptions. The changes are administrative in nature because they provide clarification and correction of existing expectations, and therefore the proposed change does not increase the probability of any accident previously evaluated. No physical or operational changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

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Enclosure 1 Basis for Proposed Changes The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, and therefore do not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no design changes associated with the proposed changes, and the changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The changes do not alter assumptions made in the safety analysis, and are consistent with the safety analysis assumptions and current plant operating practice. Due to the administrative nature of the changes, they cannot be an accident initiator.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes are administrative in nature and do not result in a change in the manner in which the plant operates. The proposed changes provide clarification and correction of existing expectations that do not degrade the availability or capability of safety related equipment, or alter their operation. The proposed changes do not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed changes will not result in plant operation in a configuration outside the design basis. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.12 TSTF-308-A, Revision 1, "Determination of Cumulative and Projected Dose Contributions in RECP" Description of Proposed Change The proposed change revises Specification 5.5.4, "Radioactive Effluent Controls Program," paragraph e, to describe the original intent of the dose projections.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification The proposed change revises Specification 5.5.4, "Radioactive Effluent Controls Program," paragraph e, to describe the original intent of the dose projections.

The NRC's draft Standard Technical Specifications for four-loop Westinghouse plants (8/14/87 letter to Texas Utilities) included Radioactive Effluent Technical Specifications. The two Surveillances in those draft Standard Technical Specifications reflect the intent of Farley Specification 5.5.4, paragraph e. SR 4.11.1.2 for Dose stated, "Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days." SR 4.11.1.3.1 for Liquid Radwaste Treatment System stated, "Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized." Generic Letter 89-01 inappropriately combined these two Surveillance Requirements for cumulative and projected doses and can be interpreted to require determining projected dose contribution for the current calendar quarter and current calendar year every 31 days. Therefore, the proposed change clarifies the wording in 5.5.4.e to not require dose projections for a calendar quarter and a calendar year every 31 days.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not issue a letter approving TSTF-308-A, Revision 1; however, it was incorporated by the NRC into Revision 2 of the ISTS NUREGs. An example of a plant-specific NRC approval of the changes in TSTF-308-A is Calvert Cliffs Units 1 and 2 Amendment Numbers 259/236 dated July 16, 2003 (ACN ML031330142).

List of Affected Pages 5.5-3 El -47

Enclosure 1 Basis for Proposed Changes Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 64, Monitoring Radioactivity Releases, states:

Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

The proposed change is an administrative requirement related to monitoring effluent discharge. It clarifies the intent of the NRC's guidance published in Generic Letter 89-01.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises Specification 5.5.4, "Radioactive Effluent Controls Program," paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are not an assumption in any accident previously evaluated and have no effect on the mitigation of any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed El -48

Enclosure 1 Basis for Proposed Changes change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises Specification 5.5.4, "Radioactive Effluent Controls Program," paragraph e, to describe the original intent of the dose projections. The cumulative and projection of doses due to liquid releases are administrative tools to assure compliance with regulatory limits, The proposed change revises the requirement to clarify the intent, thereby improving the administrative control over this process. As a result, any effect on the margin of safety should be minimal. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.13 TSTF-312-A, Revision 1, "Administrative Control of Containment Penetrations" Description of Proposed Change The proposed TS change adds a Note to the LCO for Specification 3.9.3, "Containment Penetrations," allowing penetration flow path(s) that have direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative control.

Differences Between the Proposed Change and the Approved Traveler The Farley Section 3.9 specification numbers are different from the ISTS Section 3.9 specification numbers. Farley Specification 3.9.3, "Containment Penetrations," is equivalent to Specification 3.9.4 in the ISTS. This has no effect on the requested change.

Farley LCO 3.9.3.b was previously amended to allow the personnel and equipment airlocks to remain open during core alterations or movement of irradiated fuel assemblies within the containment, provided one airlock door was available and a designated individual was available to close the open airlock door(s) if needed. The scope of this previous amendment overlaps the scope of TSTF-312-A, and as a result LCO 3.9.3 and its associated Bases differ from that presented in TSTF-312-A. The Note for LCO 3.9.3 and the supplemental LCO text for Bases 3.9.3 are incorporated without change from TSTF-312-A. No additional changes to the LCO and Bases were necessary of made as a result of the existing allowance for the personnel and equipment airlock.

Summary of the Approved Traveler Justification The proposed TS change adds a Note to the LCO for Specification 3.9.3, "Containment Penetrations," allowing "Penetration flow path(s) that have direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative control." The Applicability for LCO 3.9.3 is during core alterations, and during movement of irradiated fuel assemblies within containment.

The changes proposed in TSTF-312-A, Revision 1 are consistent with those in Specification 3.6.3, "Containment Isolation Valves." TS 3.6.3, Actions Note 1, allows penetration flow path(s) (except for the 24 inch purge valves) to be unisolated intermittently under administrative control, and is Applicable in MODES 1, 2, 3, and 4. Under the applicable conditions for LCO 3.6.3, the accident analyses credit the primary containment as a release barrier. The proposed change to LCO 3.9.3 would be Applicable under significantly lower energy conditions than those that apply for LCO 3.6.3, and is therefore less risk significant. Adoption of this change is proposed to provide a consistent approach to containment boundary issues that utilizes previously approved and acceptable compensatory measures.

The proposed change also includes the addition of text to the LCO discussion in Bases 3.9.3 stipulating that the administrative controls that are put in place when penetrations flow path(s) are unisolated ensure that: 1) appropriate personnel are aware of the open status of the penetration flow path during core alterations or El - 50

Enclosure 1 Basis for Proposed Changes movement of irradiated fuel assemblies within the containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident (FHA).

TSTF-312-A includes a Reviewer's Note that identifies the need for a confirmatory FHA dose calculation that has been accepted by the NRC staff, and that indicates acceptable radiological consequences. NRC acceptance of the Farley FHA dose calculation was documented during review of the prior license amendment request affecting LCO 3.9.3 that allowed the personnel and equipment airlocks to remain open during core alterations or movement of irradiated fuel assemblies within the containment (see LCO 3.9.4.b). NRC acceptance of this change was based on doses for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months release and a licensee commitment for a designated and available person to close the airlock door. This acceptance is documented in a letter from Sean Peters (NRC) to L. M.

Stinson (SNC), Amendments 165/157, dated September 30, 2004 (ACN ML042820368).

The Reviewer's Note identifies licensee commitments to implement administrative procedures that ensure the open containment airlock can be promptly closed in the event of an FHA following personnel evacuation, and that open penetration flow path(s) can be promptly closed. The Reviewer's Note also identifies that the time to close such penetrations, or combinations of penetrations, should be included in the confirmatory dose calculations.

The Farley FHA dose calculation analyzes offsite and control room doses for FHA events within containment, and evaluates scenarios where the equipment hatch and/or personnel airlocks are open. Doses are calculated for the 0-2 hour period, and essentially all of the activity that is released into containment by the FHA event is released from containment during these 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . There is fundamentally no contribution to calculated doses after the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the event.

Calculated doses for the inside containment FHA events, with the equipment hatch and/or personnel airlocks open, are within the Standard Review Plan criteria of 6 REM to the whole body and 75 REM to the thyroid, and are also within the control room dose acceptance criteria from 10 CFR 50, Appendix A, General Design Criterion 19, of 5 REM whole body, and the interpreted limit of 50 REM thyroid, as described in R.G. 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors."

Due to the fact that essentially all of the activity that is released into containment during the inside containment FHA event scenarios is assumed to be released during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the event, calculated doses for a release through an open equipment hatch and/or open personnel airlock doors are very nearly the same. Offsite and control room doses for the inside containment FHA event scenarios where the equipment hatch and/or personnel airlocks are open were shown to be within acceptance limits without assuming that leak paths are isolated during the scenario.

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Enclosure 1 Basis for Proposed Changes Similarly, offsite and control room doses resulting from either a simultaneous or individual release though one or more open containment penetrations, an open equipment hatch, or open personnel airlock doors will be very nearly the same as those calculated for an open equipment hatch and/or open personnel airlock doors, and will also be within acceptance limits without assuming that any leak paths are isolated. Consequently, it is not necessary to provide the time to close unisolated containment penetration(s) in the FHA dose calculations.

Allowing penetration flow paths to be unisolated during core alterations or movement of irradiated fuel will not invalidate the conclusion that the potential dose consequences from a FHA are within 10 CFR Part 100 limits.

SNC will establish administrative controls to ensure: 1) appropriate personnel are aware of the open status of the penetration flow path(s) during core alterations or movement of irradiated fuel assemblies within the containment, and 2) specified individuals are designated and readily available to isolate any open penetration flow path(s) in the event of an FHA inside containment.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None.

Licensee Commitments Required to Adopt this Change

1. Administrative controls will be established to ensure appropriate personnel are aware of the open status of the penetration flow path(s) during core alterations or movement of irradiated fuel assemblies within the containment.

2. Existing administrative controls for open containment airlock doors will be expanded to ensure specified individuals are designated and readily available to isolate any open penetration flow path(s) in the event of an FHA inside containment.

NRC Approval TSTF-312-A, Revision 1, was approved by the NRC as documented in a letter from William Beckner (NRC) to James Davis (NEI), dated August 16, 1999 (ACN 9908250220). TSTF-312-A, Revision 1, has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML0212110540). An example of a plant-specific NRC approval of the changes in TSTF-312-A, Revision 1 is Arkansas Nuclear One, Unit 2, Amendment Number 245 dated August 10, 2011 (ACN ML111940085).

List of Affected Pages 3.9.3-1 B3.9.3-4 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criteria:

Criterion 19, Control Room, states:

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Enclosure 1 Basis for Proposed Changes A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Criterion 56-Primary containment isolation, states:

Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety The proposed change to LCO 3.9.3 will allow containment penetration flow path(s) to be open during refueling operations under administrative control. This change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by the existing FHA accident analysis.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

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Enclosure 1 Basis for Proposed Changes Signification Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change would allow containment penetrations to be unisolated under administrative controls during core alterations or movement of irradiated fuel assemblies within containment. The status of containment penetration flow paths (i.e., open or closed) is not an initiator for any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not affect the design of the primary containment, or alter plant operating practices such that the probability of an accident previously evaluated would be significantly increased. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by the fuel handling accident (FHA) analysis. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Allowing penetration flow paths to be open is not an initiator for any accident. The proposed change to allow open penetration flow paths will not affect plant safety functions or plant operating practices such that a new or different accident could be created. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

TS 3.9.3 provides measures to ensure that the dose consequences of a postulated FHA inside containment are minimized. The proposed change El - 54

Enclosure 1 Basis for Proposed Changes to LCO 3.9.3 will allow penetration flow path(s) to be open during refueling operations under administrative control. These administrative controls will provide assurance that prompt closure of open penetrations flow paths can and will be achieved in the event of an FHA inside containment, and will minimize dose consequences. The proposed change is bounded by the existing FHA analysis. The proposed change does not affect the safety analysis acceptance criteria for any analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined, nor is there any adverse effect on those plant systems necessary to assure the accomplishment of protection functions. The proposed change will not result in plant operation in a configuration outside the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.14 TSTF-314-A, Revision 0, "Require Static and Transient FQ Measurement" Description of Proposed Change The proposed change revises the Required Actions of Specification 3.1.4, "Rod Group Alignment Limits," and Specification 3.2.4, "Quadrant Power Tilt Ratio," to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z).

Differences Between the Proposed Change and the Approved Traveler The Farley Section 3.1 specification numbers are different from the ISTS Section 3.1 specification numbers. Farley Specification 3.1.4, "Rod Group Alignment Limits," is equivalent to Specification 3.1.5 in the ISTS. This has no effect on the requested change.

The Bases changes associated with TSTF-314-A are adopted with exception.

TSTF-314-A modifies the Bases to state that verification that FQ(Z) is within limits requires verification of both the steady state and transient portions of FQ(Z). The Farley Specifications do not use the same symbols as the ISTS for the steady state and transient portion of FQ(Z). The Bases are revised to reflect the Farley terminology.

Summary of the Approved Traveler Justification FQ(Z) is approximated by both a steady state and transient component of FQ.

When Actions require that FQ(Z) be verified to be within limits, both the steady state and transient portions of FQ(Z) should be confirmed to be within their limits.

Currently, the Rod Group Alignment Limits and Quadrant Power Tilt Specifications only require measurement of the steady state FQ(Z), as determined by SR 3.2.1.1. Both specifications are revised to also require measurement of the transient FQ(Z), as determined by SR 3.2.1.2. This change will ensure that the hot channel factors are within their limits when the rod alignment limits or quadrant power tilt ratio are not within their limits.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-314-A, Revision 0 in a letter from William D. Beckner (NRC) to James Davis (NEI) dated January 13, 1999 (ACN 9901210038). TSTF-314-A has been adopted by many plants as part of complete conversion to the ISTS, such as Donald C. Cook Nuclear Plant Amendment Numbers 287/269, dated June 1, 2005 (ACN ML050620034).

List of Affected Pages 3.1.4-2 3.2.4-1 3.2.4-3 El - 56

Enclosure 1 Basis for Proposed Changes B3.1.4-7 B3.2.4-3 B3.2.4-5 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met. The proposed change makes the remedial actions taken when the Heat Flux Hot Channel Factor, FQ(Z), is not within its limit consistent with the LCO. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the Required Actions of Specification 3.1.4, "Rod Group Alignment Limits," and Specification 3.2.4, "Quadrant Power Tilt Ratio," to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z). This change will ensure that the hot channel factors are within their limits when the rod alignment limits or quadrant power tilt ratio are not within their limits. The verification of hot channel factors is not an initiator of any accident previously evaluated. The verification that both the steady state and transient portion of FQ(Z) are within their limits will ensure this initial assumption of the accident analysis is met should a previously evaluated accident occur. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

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Enclosure 1 Basis for Proposed Changes

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the Required Actions in the Specifications for Rod Group Alignment Limits and Quadrant Power Tilt Ratio to require measurement of both the steady state and transient portions of the Heat Flux Hot Channel Factor, FQ(Z). This change is a correction that ensures that the plant conditions are as assumed in the accident analysis.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.15 TSTF-315-A, Revision 0, "Reduce Plant Trips Due to Spurious Signals to the NIS During Physics Testing" Description of Proposed Change The proposed change revises Specification 3.1.8, "PHYSICS TESTS Exceptions

- MODE 2," to allow the number of channels required by LCO 3.3.1, "RTS Instrumentation," to be reduced from "4" to "3" to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition.

Differences Between the Proposed Change and the Approved Traveler The NUREG-1431 changes provided in TSTF-315-A are based on a four-loop Westinghouse plant design. Farley Units 1 and 2 are 3-loop plants and have only three Overtemperature AT instrument channels. The TSTF-315-A changes which allow reducing the minimum number of required channels from "4" to "3" are therefore not applicable for Farley, and are not adopted.

Additionally, the Farley Section 3.1 specification numbers are different from the ISTS Section 3.1 specification numbers. Farley Specification 3.1.8, "PHYSICS TESTS - MODE 2" is equivalent to Specification 3.1.10 in the ISTS. This has no effect on the requested change.

TSTF-315-A identifies a physics test exception to the following Specification 3.3.1 Functions:

NUREG-1431 Equivalent Farley Function Number Description Function Number Function 2 Power Range Function 2 Neutron Flux Function 3 Power Range Function 3 Neutron Flux Rate Function 18.e Reactor Trip Function 17.e System Interlocks, Power Range Neutron Flux, P-10 Therefore, the Farley LCO is modified to apply to Functions 2, 3, and 17.e.

Summary of the Approved Traveler Justification During the performance of physics testing, one power range nuclear instrumentation channel is used to provide input to the reactivity computer. When this channel is used, the channel is usually placed in a tripped condition by removing the fuses to the electronics drawer. This effectively places the reactor trip logic in a one-out-of-three logic status. Any spurious signals received on one channel will result in a reactor trip. The proposed change allows retaining the fuses in the nuclear instrumentation channel that is disconnected from the detector input. This would effectively place this channel in a bypass state and El - 59

Enclosure I Basis for Proposed Changes place the overall logic in a two-out-of-three logic status. This would preclude a spurious signal on one channel from causing a plant trip.

During the performance of Physics Tests, the reactor power is restricted to less than 5 percent. At this power level, the intermediate range neutron flux trip is active and also provides reactor trip protection. During physics testing, the plant is held in a stable state with minimal changes to steam or feed flow. In addition, the Nuclear Instrumentation System (NIS) trip setpoints are typically set to reduced values until after the core physics have been verified following a reload.

The proposed change does not eliminate the requirement for the P-10 interlock to be operable. The interlock enables the low power range and intermediate power range trips before entry into Mode 2 for the low power physics testing with power below the lower P-10 allowable values. Therefore, the low power range and intermediate power range trips are already enabled before the low power testing begins, and bypassing the instrumentation channel will not affect these low power reactor trips. As a result, placing the NIS in a two-out-of-three logic versus a one-out-of-three logic reduces the risk of unnecessary plant trips during the performance of physics tests while providing adequate reactor protection.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-315-A, Revision 0, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated June 29,1999 (ACN 9907060395). An example of a plant-specific NRC approval of the changes in TSTF-315-A is Wolf Creek Amendment Number 151 dated April 21, 2003 (ACN ML031120699).

List of Affected Pages 3.1.8-1 B3.1.8-5 Applicable Regqulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.55a(h),

"Protection and Safety Systems," requires meeting IEEE Standard 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations."

Appendix A to 10 CFR, Part 50, "General Design Criteria for Nuclear Power Plants," contains several general design criteria that are applicable to instrumentation and control, such as:

Criterion 13, "Instrumentation and Control" Criterion 20, "Protection Systems Functions" Criterion 21, "Protection System Reliability and Testability" Criterion 22, "Protection System Independence" Criterion 23, "Protection System Failure Modes" El -60

Enclosure 1 Basis for Proposed Changes Criterion 24, "Separation of Protection and Control Systems," and Criterion 25, "Protection System Requirements for Reactivity Control Malfunctions."

The proposed change does not affect the design of the plant instrumentation, but provides an operational allowance needed to safely perform required testing.

There is no regulatory requirement that specifies what operational allowances should be included in the Technical Specifications. The allowance is consistent with the safety significance of the transitory condition. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises Specification 3.1.8, "PHYSICS TESTS Exceptions - MODE 2," to allow the number of channels required by LCO 3.3.1, "RTS Instrumentation," to be reduced from "4" to "3" to allow one nuclear instrumentation channel to be used as an input to the reactivity computer for physics testing without placing the nuclear instrumentation channel in a tripped condition. A reduction in the number of required nuclear instrumentation channels is not an initiator to any accident previously evaluated. With the nuclear instrumentation channel placed in bypass instead of in trip, reactor protection is provided by the intermediate range neutron flux detectors and the nuclear instrumentation system operating in a two-out-of-three channel logic. As a result, the ability to mitigate any accident previously evaluated is not significantly affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to El - 61

Enclosure 1 Basis for Proposed Changes the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change reduces the probability of a spurious reactor trip during physics testing. The reactor trip system continues to be capable of protecting the reactor utilizing the intermediate range neutron flux reactor trip and the power range neutron flux trips operating in a two-out-of-three trip logic. As a result, the reactor is protected and the probability of a spurious reactor trip is significantly reduced. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.16 TSTF-325, Revision 0, "ECCS Conditions and Required Actions with Less Than 100% Equivalent ECCS Flow" Description of Proposed Change The proposed change revises LCO 3.5.2, "ECCS - Operating," to remove the "AND at least 100% ECCS flow equivalent to a single OPERABLE ECCS train available" from Condition A, and creates a new Condition C which states, "With less than 100% ECCS flow equivalent to a single OPERABLE ECCS train available." Required Action C.1 will require the unit to enter LCO 3.0.3 immediately.

Differences Between the Proposed Change and the Approved Traveler None.

Summary of the Approved Traveler Justification LCO 3.5.2, "ECCS - Operating," has the potential to be interpreted in a manner that would incorrectly address an inoperable Emergency Core Cooling System (ECCS). Condition A of LCO 3.5.2 addresses a situation involving "One or more trains of ECCS inoperable AND at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available." This allows inoperabilities to be present in both ECCS trains, as long as 100% equivalent flow is available; for example, high pressure injection inoperable in Train A and low pressure injection inoperable in Train B. If a situation were to occur which resulted in less than 100% ECCS flow (such as both low pressure injection pumps inoperable), then LCO 3.0.3 would be entered. However, the stated conditions for Condition A would no longer be applicable, as there was less than the 100% equivalent flow.

It could be interpreted from the "AND" that Condition A is exited when LCO 3.0.3 is entered. This is in conflict with Section 1.3 of the Technical Specifications on Completion Times, specifically Example 1.3-2. The intent is that even though LCO 3.0.3 is entered, the applicable Condition of the affected LCO (in this case, Condition A of LCO 3.5.2) should not be exited. Condition A should still be applicable, and the time tracked while in LCO 3.0.3. This will allow a smooth transition should a pump be restored and LCO 3.0.3 exited. This is accomplished by breaking up Condition A into two separate Conditions, such that with any pump or train inoperable, Condition A will still be applicable. The proposed change corrects the structure of the LCO to assure its correct application. There is no change in intent or in the way the LCO is actually applied.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-325-A, Revision 0, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated June 29, 1999 (ACN 9907060395). TSTF-325-A, Revision 0 has been adopted by many plants El - 63

Enclosure 1 Basis for Proposed Changes as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML021200265). An example of a plant-specific NRC approval of the changes in TSTF-325-A is Wolf Creek Amendment Number 182 dated May 15, 2009 (ACN ML090960734).

List of Affected Pages 3.5.2-1 B3.5.2-7 B3.5.2-8 Applicable Reaqulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 35, Emergency Core Cooling, states:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

The proposed TS changes correct the structure of LCO 3.5.2 to assure its proper application. There will be no changes to the ECCS system design, or the intent or manner in which the LCO is applied, that would cause compliance with the regulatory requirements to come into question. ECCS systems will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

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Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change corrects the structure of Technical Specification 3.5.2 to assure its proper application. There is no change in intent or in the way the Technical Specification is applied. The literal (and unintended) interpretation of the existing LCO structure could, under some circumstances, provide longer than intended Completion Times for restoration of operability. The proposed change only clarifies the requirements of the Required Actions. Since the proposed change affects neither the Technical Specification intent, nor its application, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated..

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change corrects the structure of the Technical Specification to assure its correct application. There is no change in intent or in the way the Technical Specification is applied. The proposed changes would not result in any physical alterations to the plant configuration, no new equipment is added, no equipment interfaces are modified, and no changes to any equipment's function or the method of operating the equipment are being made. As the proposed changes would not change the design, configuration or operation of the plant, no new or different kinds of accident modes are created. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change corrects the structure of the Technical Specification to assure its correct application. There is no change in intent or in the way the Technical Specification is applied. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.17 TSTF-340-A, Revision 3, "Allow 7 Day Completion Time for a Turbine-Driven AFW Pump Inoperable" Description of Proposed Change The proposed change revises Specification 3.7.5, "Auxiliary Feedwater System,"

to allow a 7 day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3 immediately following a refueling outage if Mode 2 has not been entered.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification Present specifications have a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for any inoperable Auxiliary Feedwater (AFW) pump with an Action to be in Mode 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months if the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is not met. The proposed change would allow a 7 day Completion Time for the turbine-driven AFW pump if the inoperability occurs in Mode 3 immediately following a refueling outage if Mode 2 has not been entered. This change will reduce the number of unnecessary Mode changes by providing added flexibility in Mode 3 to repair and test the turbine-driven AFW pump following a refueling outage. In the proposed condition, there is minimal decay heat due to the decay of the irradiated fuel during the refueling outage and the replacement of irradiated fuel with unirradiated fuel. The change is reasonable given the redundant capabilities afforded by the AFW system, the time needed to perform repairs and testing of the turbine-driven pump, and the low probability of an accident occurring during this time period that would require the operation of the turbine driven pump. In addition, there are alternate methods, such as feed and bleed, available to remove decay heat if necessary.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-340-A, Revision 3, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated March 16, 2000 (ACN ML003694199). An example of a plant-specific NRC approval of the changes in TSTF-340-A is Palo Verde Units 1, 2, and 3 Amendment Numbers 134/134/134 dated March 29, 2001 (ACN ML010930242).

List of Affected Pages 3.7.5-1 B3.7.5-5 B3.7.5-6 El - 66

Enclosure 1 Basis for Proposed Changes Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met. The proposed change makes the remedial actions consistent with safety significance of the condition when a turbine-driven Auxiliary Feedwater pump is inoperable and the reactor core is in a low decay-heat state following a refueling outage. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises Specification 3.7.5, "Auxiliary Feedwater (AFW) System," to allow a 7 day Completion Time to restore an inoperable turbine-driven pump in Mode 3 immediately following a refueling outage, if Mode 2 has not been entered. An inoperable AFW turbine-driven pump is not an initiator of any accident previously evaluated. The ability of the plant to mitigate an accident is no different while in the extended Completion Time than during the existing Completion Time. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

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Enclosure 1 Basis for Proposed Changes The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises Specification 3.7.5, "Auxiliary Feedwater (AFW) System," to allow a 7 day Completion Time to restore an inoperable turbine-driven AFW pump in Mode 3 immediately following a refueling outage if Mode 2 has not been entered. In Mode 3 immediately following a refueling outage, core decay heat is low and the need for AFW is also diminished. The two operable motor driven AFW pumps are available and there are alternate means of decay heat removal if needed.

As a result, the risk presented by the extended Completion Time is minimal. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.18 TSTF-343, Revision 1, "Containment Structural Integrity" Description of Proposed Change The proposed changes revise the Containment Leakage Rate Testing Program in TS Section 5.5, for consistency with the requirements of 10 CFR 50.55a(g)(4) for components classified as Code Class CC. This regulation requires licensees to update their containment inservice inspection requirements in accordance with Subsections IWE and IWL of Section Xl, Division I of the ASME Boiler and Pressure Vessel Code as limited by 10 CFR 50.55a(b)(2)(vi) and modified by 10 CFR 50.55a(b)(2)(viii) and 10 CFR 50.55a(b)(2)(ix).

The Containment Leakage Rate Testing Program description will be revised to add the following exception to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Testing Program,"

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section Xl Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

The TS Bases for SR 3.6.1.1 are also revised for consistency with the requirements of the ASME Code Section XI, Subsection IWL and applicable addenda as required by 10 CFR 50.55a.

Differences Between the Proposed Change and the Approved Traveler The Administrative Controls numbering in Farley TS Section 5.5 differs from the ISTS Administrative Controls numbering. Farley TS 5.5.17, "Containment Leakage Rate Testing Program," is equivalent to TS 5.5.16 in the ISTS. This has no effect on the requested change.

The changes identified in TSTF-343-A, Revision 1, for the TS 5.5.6, "Containment Tendon Surveillance Program," and conforming changes to the Bases for SR 3.6.1.2 and the TS 3.6.1 Bases References, are not adopted. The changes in TSTF-343-A that affect this program are already reflected in the Farley Technical Specifications, and are therefore not necessary.

Additionally, the Farley Containment Leakage Rate Testing Program, as described in TS 5.5.17, contains the following exception:

Section 9.2.3: The next Type A test, after the March 1994 test for Unit 1 and the March 1995 test for Unit 2, shall be performed during refueling outage R22 (Spring 2009) for Unit 1 and during El - 69

Enclosure 1 Basis for Proposed Changes refueling outage R20 (Spring 2010) for Unit 2. This is a one-time exception.

This exception provided a one-time exception for actions that have now been performed. Retention of this exception in the TS for historical purposes is not necessary, and the exception is therefore deleted from TS 5.5.17.

Summary of the Approved Traveler Justification On January 7, 1994, the Nuclear Regulatory Commission (NRC) published a proposed amendment to the regulations to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of Section XI, Division I of the ASME Boiler and Pressure Vessel Code (the Code). The final rule, Subpart 50.55a(g)(6)(ii)(B) of Title 10 of the Code of Federal Regulations (10 CFR), became effective on September 9, 1996, and requires licensees to implement Subsections IWE and IWL, with specified modifications and limitations, by September 9, 2001.

The Farley containment consists of a prestressed, reinforced concrete cylindrical structure with a shallow domed roof and a reinforced concrete foundation slab.

The cylindrical portion of the containment is prestressed by a post-tensioning system composed of horizontal and vertical tendons. A 1/4-in.-thick welded steel liner is attached to the inside face of the concrete. The floor liner is installed on top of the foundation slab and covered with concrete.

TS Section 5.5.17 requirements for the Containment Leakage Rate Testing Program specify that the program shall be in accordance with the guidelines contained in RG 1.163. Regulatory Position C.3 of this regulatory guide states:

Section 9.2.1, "Pretest Inspection and Test Methodology," of NEI 94-01 provides guidance for the visual examination of accessible interior and exterior surfaces of the containment system for structural problems.

These examinations should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test if the interval, for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration.

There are no specific requirements in NEI 94-01 for the visual examination except that it is to be a general visual examination of accessible interior and exterior surfaces of the primary containment components.

In addition to the requirements of RG 1.163 and NEI 94-01, the concrete surfaces of the containment must be visually examined in accordance with the ASME Section XI Code, Subsection IWL, and the liner plate inside containment must be visually examined in accordance with Subsection IWE. The frequency of visual examination of the concrete surfaces per Subsection IWL is once every five years alternating between units for a two-unit plant, and the frequency of visual examination of the liner plate per Subsection IWE is, in general, three visual examinations over a 10-year period. The visual examinations performed pursuant to Subsection IWL may be performed at any time during power operation or during shutdown, and the visual examinations performed pursuant to El - 70

Enclosure 1 Basis for Proposed Changes Subsection IWE are performed during refueling outages since this is the only time that the liner plate is fully accessible.

The visual examinations performed pursuant to Subsections IWL and IWE are more rigorous than those performed pursuant to RG 1.163 and NEI 94-01. For example, Subarticle IWE-2320 requires the general visual examination to be the responsibility of an individual who is knowledgeable in the requirements for design, inservice inspection, and testing of Class MC and metallic liners of Class CC components. Subsection IWE, Subarticle-2330 requires the examination to be performed either directly or remotely, by an examiner with visual acuity sufficient to detect evidence of degradation. Subarticle IWL-2320 requires the responsible engineer to be a Registered Professional Engineer and provides minimum knowledge and experience requirements. Furthermore, visual examinations of both the concrete surfaces and the liner plate must be reviewed by an Inspector employed by a State or municipality of the United States or an Inspector regularly employed by an insurance company authorized to write boiler and pressure vessel insurance, in accordance with IWA-21 10 and IWA-2120.

The combination of the Code requirements for the rigor of the visual examinations plus the third party review more than offsets the fact that fewer visual examinations of the concrete will be performed during a 10-year interval. The fact that the concrete visual examination pursuant to Subsection IWL may be performed during power operation as well as during a refueling outage will have no effect on the quality of the examination and will provide flexibility in scheduling of the visual examinations.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-343-A, Revision 1, in a letter from Thomas H. Boyce (NRC) to the Technical Specification Task Force, dated December 6, 2005 (ACN ML053460302). An example of a plant-specific NRC approval of the changes in TSTF-340-A is Diablo Canyon, Units 1 and 2, Amendment Numbers 197/198, dated June 26, 2007 (ACN ML071370731).

List of Affected Pages 5.5-14 B3.6.1-4 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

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Enclosure 1 Basis for Proposed Changes Criterion 16, Containment Design, states:

Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

Appendix J to 10 CFR, Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," contains the following pertinent criterion:

Option B, Performance Based Requirements, states:

A general inspection of the accessible interior and exterior surfaces of the containment structures and components shall be performed prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leaktightness.

When implementing Option B of Appendix J to 10 CFR Part 50, "Performance-Based Leakage-Test Requirements," TS 5.5.17 states that the licensee's program shall be in accordance with the guidelines in RG 1.163, "Performance-Based Containment Leak-Test Program." Regulatory Position C.3 of RG 1.163 discusses visual examinations of accessible interior and exterior surfaces of the containment system. Specifically, Regulatory Position C.3 states, "examinations should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years..."

In addition, Section 50.55a of 10 CFR Part 50 requires licensees to perform their containment ISI requirements in accordance with Subsections IWE and IWL of Section XI, Division I of the ASME Code. Paragraph 50.55a(g)(4) of 10 CFR requires licensees to update their containment ISI requirements in accordance with subsections IWE and IWL of Section XI, Division I, of the ASME Code as limited by 10 CFR 50.55a(b)(2)(vi) and modified by 10 CFR 50.55a(b)(2)(vii) and 10 CFR 50.55a(b)(2)(ix).

The proposed TS changes revise the containment leakage rate testing program for consistency with the requirements of 10 CFR 50.55a(g)(4) for components classified as ASME Code CC. Specifically, TS Section 5.5.17, "Containment Leakage Rate Testing Program," is revised to allow the performance of visual examinations of the containment pursuant to ASME Code, Section XI, Subsections IWE and IWL, in lieu of the visual examinations performed pursuant to RG 1.163.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

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Enclosure 1 Basis for Proposed Changes Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes revise the Technical Specifications (TS)

Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage.

The frequency of visual examinations of the containment and the mode of operation during which those examinations are performed does not affect the initiation of any accident previously evaluated. The use of NRC approved methods and frequencies for performing the inspections will ensure the containment continues to perform the mitigating function assumed for accidents previously evaluated. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes revise the TS Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage.

The proposed changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure. Therefore, the proposed El - 73

Enclosure 1 Basis for Proposed Changes changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes revise the Technical Specifications (TS)

Administrative Controls programs for consistency with the requirements of 10 CFR 50, paragraph 55a(g)(4) for components classified as Code Class CC. The proposed changes affect the frequency of visual examinations that will be performed for the concrete surfaces of the containment for the purpose of the Containment Leakage Rate Testing Program, and allows those examinations to be performed during power operation in addition to during a refueling outage. The safety function of the containment as a fission product barrier will be maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.19 TSTF-349-A, Revision 1, "Add Note to LCO 3.9.5 Allowing Shutdown Cooling Loops Removal from Operation" Description of Proposed Change The proposed change adds an LCO Note to LCO 3.9.5, "RHR and Coolant Circulation - Low Water Level," to allow the securing of the operating train of RHR for up to 15 minutes to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees Fahrenheit below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration below the Technical Specifications limit are prohibited.

Differences Between the Proposed Change and the Approved Traveler The Farley Specifications in Section 3.9, "Refueling Operations," are numbered differently than the equivalent specifications in the ISTS. ISTS Specification 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level," is equivalent to Farley Specification 3.9.5. This has no effect on the requested change.

Summary of the Approved Traveler Justification The proposed change adds an LCO Note to LCO 3.9.5, "RHR and Coolant Circulation - Low Water Level," to allow the securing of the operating train of RHR to support switching operating trains. The allowance is acceptable because the allowed time frame is short and limitations are in place to ensure the RCS boron concentration limit is not reduced and to preclude draining activities. The proposed Note is consistent with the allowance LCO 3.4.8, "RCS Loops - MODE 5, Loops not filled." With the plant in MODE 6 with less than 23 feet of water above the Reactor Vessel flange, the reactor coolant system (RCS) is in an inventory status similar to LCO 3.4.8. Therefore, the allowances should also be consistent.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC did not issue a letter approving TSTF-349-A, Revision 1; however, it was incorporated by the NRC into Revision 2 of the ISTS. An example of a plant-specific NRC approval of the changes in TSTF-349-A is Calvert Cliffs Units 1 and 2 Amendment Numbers 256/233 dated February 25, 2003 (ACN ML030560015).

List of Affected Pages 3.9.5-1 B3.9.5-2 El - 75

Enclosure I Basis for Proposed Changes Applicable Requlatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met. The proposed change makes the remedial actions consistent with safety significance of the transitory condition of swapping operating Residual Heat Removal pumps. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Si-qnificant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change adds an LCO Note to LCO 3.9.5, "RHR and Coolant Circulation - Low Water Level," to allow securing the operating train of Residual Heat Removal (RHR) for up to 15 minutes to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. Securing an RHR train to facilitate the changing of the operating train is not an initiator to any accident previously evaluated. The restrictions on the use of the allowance ensure that an RHR train will not be needed during the 15 minute period to mitigate any accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

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Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change adds an LCO Note to LCO 3.9.5, "RHR and Coolant Circulation - Low Water Level," to allow securing the operating train of RHR to support switching operating trains. The allowance is restricted to conditions in which core outlet temperature is maintained at least 10 degrees F below the saturation temperature, when there are no draining operations, and when operations that could reduce the reactor coolant system (RCS) boron concentration are prohibited. With these restrictions, combined with the short time frame allowed to swap operating RHR trains and the ability to start an operating RHR train if needed, the occurrence of an event that would require immediate operation of an RHR train is extremely remote. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.20 TSTF-355-A, Revision 0, "Changes to RTS and ESF Tables" Description of Proposed Change The proposed change modifies TS Table 3.3.1-1, "Reactor Trip System Instrumentation," and Table 3.3.2-1, "Engineered Safety Feature Actuation Instrumentation," to replace the requirement for a Trip Setpoint with a requirement for a Nominal Trip Setpoint. Additionally, a note is added to the Bases which allows: 1) the actual trip setpoint to be set more conservative than the Nominal Trip Setpoint specified in Table 3.3.1-1 or Table 3.3.2-1 in response to plant conditions, and 2) states an "as-found" trip setpoint is operable when it is outside the calibration tolerance band if the as-found value has not exceeded the associated Allowable Value and the channel is readjusted to within the established calibration tolerances.

The Bases discussion is revised to provide a conforming discussion of the LCO changes and to more clearly and accurately discuss the relation between the nominal trip setpoint, the allowable value, and the plant approved setpoint methodology. Also, the Allowable Value is clarified to be the Limiting Safety System Setting required by 10 CFR 50.36.

Differences Between the Proposed Change and the Approved Traveler The Reactor Trip System (RTS) and Engineered Safety Feature Actuations System (ESFAS) functions and notes identified in TSTF-355-A for ISTS Table 3.3.1-1 and Table 3.3.2-1 are not identical to the RTS and ESAFAS functions identified in the Farley Technical Specifications. As a result, several of the changes approved in TSTF-355-A are not applicable to Farley. The following tables summarize the differences between the TSTF-355-A proposed changes to the Westinghouse plant ISTS markups and the changes proposed for the Farley Technical Specifications:

Table 3.3.1 Reactor Trip System Instrumentation TSTF-355-A Affected RTS Function Equivalent Farley RTS Function 2.a Power Range Neutron Flux, 2.a - Adopt as proposed in TSTF-355-A.

High 2.b Power Range Neutron Flux, 2.b - Adopt as proposed in TSTF-355-A.

Low 3.a Power Range Neutron Flux Rate, 3.a - Adopt as proposed in TSTF-355-A.

High Positive Rate 3.b Power Range Neutron Flux Rate, Not applicable. Farley TS Table 3.3.1-1 High Negative Rate does not list an equivalent function.

4. Intermediate Range Neutron Flux, 4. - Adopt as proposed in TSTF-355-A.

Modes 1(c), 2 (d) (Modes 1(c), 2(d))

4. Intermediate Range Neutron Flux, 4. - Adopt as proposed in TSTF-355-A.

Mode 2(e) (Mode 2(e))

5. Source Range Neutron Flux, 5. - Adopt as proposed in TSTF-355-A.

Mode 2(e) (Mode 2(d))

(continued)

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Enclosure 1 Basis for Proposed Changes Table 3.3.1 Reactor Trip System Instrumentation (cont'd)

TSTF-355-A Affected RTS Function Equivalent Farley RTS Function

5. Source Range Neutron Flux, 5. - Adopt as proposed in TSTF-355-A.

Modes 3 (b) 4 (b), 5 (b) (Modes 3(a), 4(a ), 5(a))

8.a Pressurizer Pressure, 8.a - Adopt as proposed in TSTF-355-A.

Low 8.b Pressurizer Pressure, 8.b - Adopt as proposed in TSTF-355-A.

High 9 Pressurizer Water Level, 9. - Adopt as proposed in TSTF-355-A.

High 1O.a Reactor Coolant Flow, 10. - Adopt as proposed in TSTF-355-A.

Low (single loop) 10.b Reactor Coolant Flow, 10. - Adopt as proposed in TSTF-355-A.

Low (two loop)

12. Undervoltage RCPs 12. - Adopt as proposed in TSTF-355-A.

13. Underfrequency RCPs 13. - Adopt as proposed in TSTF-355-A.

14. Steam Generator (SG) Water 14. - Adopt as proposed in TSTF-355-A.

Level, Low-Low

15. Steam Generator (SG) Water Not applicable. Farley TS Table 3.3.1-1 Level, does not list an equivalent function.

Low

15. Steam Generator (SG) Water Not applicable. Farley TS Table 3.3.1-1 Level, does not list an equivalent function.

Low (coincident with steam flow/feedwater flow mismatch) 16.a Turbine Trip 15.a - Adopt as proposed in TSTF-355-A.

Low Auto Stop Oil Pressure 16.b Turbine Trip Not applicable. The equivalent Farley Turbine Throttle Valve Closure function (16.b) does not have a numerical trip setpoint expressed as an inequality.

18.a Intermediate Range Neutron Flux, 17.a - Adopt as proposed in TSTF-355-A.

P-6 18.c Power Range Neutron Flux, 17.c - Adopt as proposed in TSTF-355-A.

P-8 18.d Power Range Neutron Flux, 17.d - Adopt as proposed in TSTF-355-A.

P-9 18.e Power Range Neutron Flux, 17.e - Adopt as proposed in TSTF-355-A.

P-10 18.f Turbine Impulse Pressure, 17.f - Adopt as proposed in TSTF-355-A.

P-13 El - 79

Enclosure 1 Basis for Proposed Changes Table 3.3.2 Engineered Safety Feature Actuation System Instrumentation TSTF-355-A Affected ESFAS Function Equivalent Farley ESFAS Function 1 .c Safety Injection, Containment I.c - Adopt as proposed in TSTF-355-A.

Pressure, High 1 1.d Safety Injection, Pressurizer 1.d - Adopt as proposed in TSTF-355-A.

Pressure-- Low 1.e Safety Injection, Steam Line 1.e.(1) - Adopt as proposed in TSTF-355-Pressure, (1) Low A.

1.e Safety Injection, Steam Line I.e.(2) - Adopt as proposed in TSTF-355-Pressure, A.

(2) High Differential Pressure Between Steam Lines 1.f Safety Injection, High Steam Flow in Not applicable. Farley TS Table 3.3.2-1 Two Lines, Coincident with Tavg - does not include an equivalent function.

Low Low 1.g Safety Injection, High Steam Flow in Not applicable. Farley TS Table 3.3.2-1 Two Lines, Coincident with Steam does not include an equivalent function.

Line Pressure - Low 2.c Containment Spray, Containment 2.c - Adopt as proposed in TSTF-355-A.

Pressure, High - 3 3.b Containment Isolation, Phase B 3.b.(3) - Adopt as proposed in TSTF-355-Isolation, A.

(3) Containment Pressure, High - 3 4.c Steam Line Isolation, Containment 4.c - Adopt as proposed in TSTF-355-A.

Pressure, High - 2 4.d Steam Line Isolation, Steam Line 4.d - Adopt as proposed in TSTF-355-A.

Pressure, (1) Low 4.d Steam Line Isolation, Steam Line Not applicable. Farley TS Table 3.3.2-1 Pressure, does not include an equivalent function.

(2) Negative Rate -- High 4.e Steam Line Isolation, High Steam 4.e - Adopt as proposed in TSTF-355-A.

Flow in Two Steam Lines, Coincident with Tavg - Low Low 4.f Steam Line Isolation, High Steam Not applicable. Farley TS Table 3.3.2-1 Flow in Two Steam Lines, Coincident does not include an equivalent function.

with Steam Line Pressure -- Low 4.g Steam Line Isolation, High Steam Not applicable. Farley TS Table 3.3.2-1 Flow, Coincident with Safety Injection does not include an equivalent function.

and Tavg - Low Low 4.h Steam Line Isolation, High High Not applicable. Farley TS Table 3.3.2-1 Steam Flow, Coincident with Safety does not include an equivalent function.

Injection 5.b Turbine Trip and Feedwater Isolation, 5.b - Adopt as proposed in TSTF-355-A.

SG Water Level, High High (P-14)

(continued)

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Enclosure 1 Basis for Proposed Changes Table 3.3.2 Engineered Safety Feature Actuation System Instrumentation (cont'd)

TSTF-355-A Affected ESFAS Function Equivalent Farley ESFAS Function 6.c Auxiliary Feedwater, SG Water 6.b - Adopt as proposed in TSTF-355-A.

Level, Low Low 6.e Auxiliary Feedwater, Loss of Offsite Not applicable. Farley TS Table 3.3.2-1 Power does not include an equivalent function.

6.f Auxiliary Feedwater, Undervoltage 6.d - Adopt as proposed in TSTF-355-A.

Reactor Coolant Pump 6.g Auxiliary Feedwater, Trip of all Main Not applicable. The equivalent Farley Feedwater Pumps function (6.e) does not have a numerical trip setpoint expressed as an inequality.

6.h Auxiliary Feedwater, Auxiliary Not applicable. Farley TS Table 3.3.2-1 Feedwater Pump Suction Transfer on does not include an equivalent function.

Suction Pressure, Low 7.b Automatic Switchover to Not applicable. Farley TS Table 3.3.2-1 Containment Sump, Reactor Water does not include an equivalent function.

Storage Tank (RWST) Level Low -

Low, Coincident with Safety Injection 7.c Automatic Switchover to Not applicable. Farley TS Table 3.3.2-1 Containment Sump, RWST Level -- does not include an equivalent function.

Low - Low, Coincident with Safety Injection and Containment Sump Level -- High 8.b ESFAS Interlocks, Pressurizer 7.c - Adopt as proposed in TSTF-355-A.

Pressure, P-11 8.c ESFAS Interlocks, Tavg - Low Low, P- 7.d - Adopt as proposed in TSTF-355-A.

12 (increasing and decreasing values)

Available copies of TSTF-355-A do not include a change markup for page 1 of ISTS Table 3.3.2-1. In NUREG-1431, Revision 1, page 1 of Table 3.3.2-1 includes four ESFAS functions with trip setpoints that are expressed as inequalities. These functions are: 1.c (Safety Injection, Containment Pressure -

High 1), 1.d (Safety Injection, Pressurizer Pressure - Low), 1.e.(2) (Safety Injection, Steam Line Pressure, High Differential Pressure Between Steam Lines); and 1.f (Safety Injection, High Steam Flow in Two Lines, Coincident with Tavg - Low Low). The concerns involving trip setpoints that are expressed as inequalities apply to these omitted functions just as they do for the other functions that were identified in TSTF-355-A. The omitted functions were modified to remove the inequalities form their nominal trip setpoints when TSTF-355-A was incorporated into Revision 2 of the ISTS. As summarized in the table above, the nominal trip setpoints for the omitted functions have been modified, as applicable, to be consistent with the intent of the Traveler and the ISTS.

Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," is cited as a reference in the Traveler insert to the Bases El - 81

Enclosure 1 Basis for Proposed Changes Background for TS 3.3.1, but its addition to the References section of the Bases is not identified as a change in TSTF-355-A. To correct this oversight, Regulatory Guide 1.105 has been added as a reference in the Bases for TS 3.3.1.

Additionally, Regulatory Guide 1.105 is identified as an addition to the References section of the Bases for TS 3.3.2, but is not cited in the Bases text.

The addition of Regulatory Guide 1.105 to the references for Bases 3.3.2 is therefore not adopted.

Summary of the Approved Traveler Justification TSTF-355 addresses a generic NRC concern with the Technical Specifications for RTS and ESFAS instrumentation functions that are structured to reflect the ISTS prior to Revision 2. The concern is related to an observed practice involving setting of RTS and ESFAS Trip Setpoints in a manner inconsistent with the trip setpoint inequalities, with tolerances beyond the maximum and minimum (inequalities) trip setpoint values shown in the TS. This practice can render the instrument inoperable based on the ITS surveillance requirements, Limiting Conditions for Operation (LCOs), and Actions. The inequalities on the ISTS Trip Setpoints were being interpreted as limits that, when exceeded, would require entry into the appropriate LCO Action.

The proposed change revises the trip setpoint column of the RTS and ESFAS instrumentation tables to utilize a nominal trip setpoint value. Additionally, notes are provided in the Bases to clarify how the nominal trip setpoints are to be applied in the field; the relationship between the nominal trip setpoint, the allowable value, and the plant approved setpoint methodology; and how the Allowable Value relates to the Limiting Safety System Setting.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification None Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their approval of TSTF-355-A, Revision 0, in a letter from William D. Beckner (NRC) to James Davis (NEI), dated March 1, 2000 (ACN ML003684647). TSTF-355-A, Revision 0 has been adopted by many plants as part of complete conversion to the ISTS, such as North Anna Power Station (ACN ML021200265). An example of a plant-specific NRC approval of the changes in TSTF-355-A is Ginna Nuclear Power Plant, Amendment Number 85, dated September 22, 2004 (ACN ML041180293).

List of Affected Pages 3.3.1-14 3.3.1-15 3.3.1-16 3.3.1-17 3.3.1-18 3.3.1-19 El - 82

Enclosure 1 Basis for Proposed Changes 3.3.1-20 3.3.1-21 3.3.2-9 3.3.2-10 3.3.2-11 3.3.2-12 B3.3. 1-1 B3.3.1-2 B3.3.1-4 B3.3.1-5 B3.3.1-7 B3.3.1-61 B3.3.2-1 B3.3.2-3 B3.3.2-4 B3.3.2-5 Applicable Regulatory Requirements/Criteria Appendix A to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criterion:

Criterion 10, Reactor Design, states:

The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Criterion 20, Protection System Functions, states:

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are.not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(1)(ii)(A) specifies that: "... Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. ..." As such, 10 CFR 50.36 requires that limits for instrument channels that initiate protective functions must be included in the TSs.

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 10 CFR 50.36(c)(2)(ii)(C), states:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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Enclosure 1 Basis for Proposed Changes Regulatory Guide 1.105: Section C.3, interprets 10 CFR 50.36 as requiring that the limiting safety system setting (LSSS) are maintained as a TS limit.

The proposed TS changes modify Table 3.3.1-1 and Table 3.3.2-1, and their associated Bases to assure correct interpretation and application of requirements for nominal trip setpoint and LSSS values. There will be no changes to the RTS or ESFAS instrumentation design, or the intent or manner in which the LCO is applied, that would cause compliance with the regulatory requirements and guidance document above to come into question. RTS and ESFAS instrumentation and protection systems will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The RTS and ESFAS instrument functions are part of the accident mitigation response and are not themselves an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly affected by the proposed changes. The changes ensure that automatic protective actions will be initiated at or before the condition assumed in the safety analysis, and are in accordance with the intent of the Technical Specifications. The proposed changes will not cause any design or analysis acceptance criteria to be exceeded. Since there will be no adverse effect on the trip setpoints or the instrumentation associated with the trip setpoints, there will be no significant increase in the consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes include modifications to the format of the nominal trip setpoints that preserve safety analysis assumptions related to El - 84

Enclosure 1 Basis for Proposed Changes accident mitigation. The protection system will continue to initiate the protective actions as assumed in the safety analysis. The proposed changes will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis.

As the proposed changes do not change the design, configuration or operation of the plant, no new or different kinds of accident modes are created. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter any nominal trip setpoints, allowable values, or limiting safety system settings, and will continue to ensure that the trip setpoints are maintained consistent with the setpoint methodology and the plant safety analysis. The response of protection systems to accident transients reported in the Final Safety Analysis Report is unaffected by this change, and accident analysis acceptance criteria are consequently not affected. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.21 TSTF-371-A, Revision 1, "NIS Power Range Channel Daily SR TS Change to Address Low Power Decalibration" Description of Proposed Change Specification 3.3.1, "RTS Instrumentation," Surveillance 3.3.1.2 is revised to move the requirement to adjust the power range channels if the calorimetric calculated power exceeds the power range indicated power by more than +2% of Rated Thermal Power (RTP) from a Surveillance Note to the Surveillance itself.

Surveillance 3.3.1.3 is revised to move the requirement to adjust the Nuclear Instrumentation System (NIS) channel if the absolute difference between the incore detector measurement of Axial Flux Difference (AFD) and the NIS AFD indication is greater than or equal to 3% from a Surveillance Note to the Surveillance itself.

Differences Between the Proposed Change and the Approved Traveler The proposed change to the Farley Technical Specifications is editorial, not technical. The Farley Technical Specifications already contain the technical change proposed in TSTF-371-A. This technical change was approved by the NRC in Farley License Amendment Number 144/135, dated October 1, 1999. It is noteworthy that the Farley amendment is referenced in the justification of TSTF-371-A. However, the generic Traveler addressing the technical issue was not approved by the NRC until April 2, 2002. As a result, the presentation in the Farley Technical Specifications is different from the generic presentation. The proposed change to the Farley Technical Specifications revises the presentation to be consistent with the ISTS and the other SNC fleet plants.

The Bases changes proposed in TSTF-371-A and the Reviewer's Note requiring a plant specific evaluation to determine the power level below which the power range channel adjustments in a decreasing direction become a concern have already been implemented at Farley. Only the Bases changes which are needed to reflect the revised presentation are implemented.

Summary of the Approved Traveler Justification Westinghouse Technical Bulletin ESBU-TB-92-14-R1, "Decalibration Effects Of Calorimetric Power Measurements on the NIS High Power Reactor Trip at Power Levels Less Than 70% RTP," dated February 6, 1996, identified potential effects of decalibrating the NIS Power Range channels at part power operation. The decalibration may occur due to the increased uncertainty of the secondary side power calorimetric when performed at part power (less than approximately 70%

RTP). When NIS channel indication is reduced to match calculated power, the decalibration results in a non-conservative bias. The proposed change to the Technical Specifications removes the requirement to adjust the NIS instrument channel output when the indicated power is greater than the calorimetric heat balance calculation by an absolute difference of > 2% RTP. TSTF-371-A was submitted to obtain NRC approval to relax the present Technical Specifications requirements to always adjust NIS channels when indicated power differs from calorimetric heat balance calculated power by more than 2%. This Technical Specification change was submitted by Southern Nuclear Operating Company (SNC) for Farley Units 1 and 2 on November 6, 1998. The NRC approved the El - 86

Enclosure 1 Basis for Proposed Changes proposed change for Farley in License Amendment No. 144 (Unit 1)/135 (Unit 2),

dated October 1, 1999.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification The Traveler justification for the technical change to require adjustment of the NIS channels when the calculated calorimetric power exceeds the NIS channel output by +2% RTP instead of when the absolute difference between them exceeds 2% is not applicable to the proposed change to the Farley Technical Specifications.

The proposed change to the Farley Technical Specifications is editorial. Existing SR 3.3.1.2, Note 1, is moved to the body of the Surveillance. SR 3.3.1.2 states, "Compare results of calorimetric heat balance calculation to Nuclear Instrumentation System (NIS) channel output." SR 3.3.1.2 Note 1, states, "Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than +2% RTP." The proposed change deletes Note 1 and revises SR 3.3.1.2 to state, "Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP." The term "power range channel" was chosen instead of "NIS channel" as it is more precise.

Surveillance 3.3.1.3 is revised to move a requirement from the Note to the Surveillance to adjust the NIS channel if the difference between the AFD as determined by the NIS and the AFD as determined using the incore detectors exceeds 3%. This editorial change is made to be consistent with the presentation in SR 3.3.1.2.

Licensee Commitments Required to Adopt this Change None NRC Approval The NRC documented their acceptance of TSTF-371-A, Revision 1, in a Safety Evaluation transmitted by letter from William Beckner (NRC) to Anthony Pietrangelo (NEI) dated April 2, 2002 (ACN ML020940096). An example of a plant-specific approval of TSTF-371-A is Vogtle Amendment Numbers 131/110 dated April 1, 2004 (ACN ML040480089).

List of Affected Pages 3.3.1-9 B3.3.1-50 B3.3.1-51 B3.3.1-52 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(3), states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is El - 87

Enclosure 1 Basis for Proposed Changes maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

There is no regulatory requirement that specifies how Surveillance Requirements will be presented in the Technical Specifications. The proposed change makes an editorial change to an existing Surveillance Requirement. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises Specification 3.3.1, "RTS Instrumentation,"

Surveillances 3.3.1.2 and 3.3.1.3 to move requirements currently in a Note to the Surveillance itself. The change in presentation is editorial and does not affect the application of the Surveillances. The proposed change does not affect any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

El - 88

Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change revises Specification 3.3.1, "RTS Instrumentation,"

Surveillances 3.3.1.2 and 3.3.1.3 to move requirements currently in a Note to the Surveillance itself. The proposed change represents an editorial preference and does not affect the performance of the Surveillance or plant operation. The safety function tested by the Surveillance is unaffected. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

El - 89

Enclosure 1 Basis for Proposed Changes 2.22 TSTF-439-A, Revision 2, "Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO" Description of Proposed Change Specifications 3.6.6, "Containment Spray and Cooling Systems;" 3.7.5, "AFW System;" 3.8.1, "AC Sources - Operating;" and 3.8.9, "Distribution Systems -

Operating," contain Required Actions with a second Completion Time to establish a limit on the maximum time allowed for any combination of Conditions that result in a single continuous failure to meet the LCO. These Completion Times (henceforth referred to as "second Completion Times") are joined by an "AND" logical connector to the Condition-specific Completion Time and state "X days from discovery of failure to meet the LCO" (where "X" varies by specification).

The proposed change deletes these second Completion Times from the affected Required Actions. It also revises ISTS Example 1.3-3 to remove the discussion of second Completion Times and to revise the discussion in that Example to state that alternating between Conditions in such a manner that operation could continue indefinitely without restoring systems to meet the LCO is inconsistent with the basis of the Completion Times and is inappropriate. Therefore, the licensee shall have administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO.

Differences Between the Proposed Change and the Approved Traveler None Summary of the Approved Traveler Justification The proposed change adopts a new technical specification convention to limit the maximum time allowed for any combination of LCO Conditions that could result in a single continuous failure to meet the LCO. In the current Technical Specifications, a second Completion Time was included for certain Required Actions to establish a limit on the maximum time allowed for any combination of Conditions that would result in a single continuous failure to meet the LCO. In practice, the addition of second Completion Times did not create an operational restriction because the likelihood of experiencing concurrent failures such that the second Completion Time was limiting is remote. It is important to note that this issue of "flip-flopping" between Conditions only applies ifthe LCO is not met on a continuous basis. In addition, ifthe LCO requirements are met, even if for an instant, this issue does not occur.

The second Completion Times created a problem when the industry and the NRC developed and approved Technical Specification changes that integrated risk-informed Completion Times into specifications containing a second Completion Time. The problem results from extending the second Completion Time by the same amount (i.e., the second Completion Time continued to be the sum of the two Completion Times). The NRC staff expressed concerns that the extension of the second Completion Time was inappropriate because one of the two Completion Times added to obtain the second Completion Time limit was risk-based and the other was derived in a deterministic evaluation. The NRC eventually accepted the practice of adding the deterministic and risk-informed Completion Times, but it continues to result in confusion.

El - 90

Enclosure 1 Basis for Proposed Changes An alternative approach was proposed and accepted by the NRC that eliminated the second Completion Times and modified Section 1.3 of the Technical Specifications to establish a convention prohibiting alternating between Conditions, in such a manner that, operation could continue indefinitely without ever restoring systems to meet the LCO. Thus, there is no longer a specific limit to the maximum time allowed for any combination of Conditions that results in a single continuous occurrence of failing to meet the LCO.

The proposed change is appropriate because multiple continuous entries into Conditions, without meeting the LCO, will be controlled by licensee's configuration risk management programs, which were implemented to meet the requirements 10 CFR 50.65 (the Maintenance Rule) to assess and manage risk, and controlled by the Use and Application convention discussed in Section 1.3 of the Technical Specifications. These controls provide adequate assurance against inappropriate use of combinations of Conditions that result in a single contiguous occurrence of failing to meet the LCO.

Differences Between the Plant-Specific Justification and the Approved Traveler Justification TSTF-439-A revises Example 1.3-3 of the Technical Specifications to state, in part, "there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended." The NRC has asked licensees adopting TSTF-439-A to provide the location of these administrative controls. As described in the section, "Licensee Commitments Required to Adopt this Change," SNC commits to revise Operations procedure FNP-0-SOP-0.13 to include these administrative controls.

The TSTF-439 justification references Regulatory Guide (RG) 1.182. On November 27, 2012, the NRC published a Federal Register Notice stating that RG 1.182 has been withdrawn and the subject matter has been incorporated into RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." RG 1.160 endorses NUMARC 93 01, Revision 4A, dated April 2011.

SNC assesses the risk of maintenance activities consistent with the guidance in RG 1.160. Therefore, the justification for adoption of TSTF-439 continues to be applicable.

Licensee Commitments Required to Adopt this Change SNC commits to revise Operations procedure FNP-0-SOP-0.13 to include a statement similar to the following: "Alternating between LCO Conditions, in order to allow indefinite continued operation while not meeting the LCO, is not allowed."

This procedure will be revised prior to implementation of the proposed change.

NRC Approval The NRC documented their approval of TSTF-439-A, Revision 2, in a letter from Thomas Boyce (NRC) to the Technical Specification Task Force date January 11, 2006 (ACN ML060120272). TSTF-439-A, Revision 2, was also incorporated in Revision 3.1 of the ISTS NUREGS. An example of a plant-specific NRC approval El - 91

Enclosure 1 Basis for Proposed Changes of the changes in TSTF-439-A is the Calvert Cliffs Amendment Numbers 304 and 282 dated January 29, 2014 (ACN ML14009A320).

List of Affected Pages 1.3-2 1.3-6 1.3-7 3.6.6-1 3.7.5-1 3.8.1-2 3.8.1-3 3.8.9-1 B3.6.6-6 B3.6.6-7 B3.7.5-6 B3.8.1-8 B3.8.1-11 B3.8.9-4 B3.8.9-5 B3.8.9-6 B3.8.9-8 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met. The proposed change makes the remedial actions consistent with likelihood and safety significance of the condition. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

El - 92

Enclosure 1 Basis for Proposed Changes

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change eliminates certain Completion Times from the Technical Specifications. Completion Times are not an initiator to any accident previously evaluated. As a result, the probability of an accident previously evaluated is not affected. The consequences of an accident during the remaining Completion Time are no different than the consequences of the same accident during the removed Completion Times. Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change to delete the second Completion Time does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed changes will not result in plant operation in a configuration outside of the design basis. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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Enclosure 1 Basis for Proposed Changes 2.23 ISTS Adoption #1 - Revise LCO 3.3.2 ESFAS Interlock P-4 Required Action Completion Time Description of Proposed Change Table 3.3.2-1, "Engineered Safety Feature Actuation System Instrumentation,"

Function 7b, "Reactor Trip, P-4," is revised to require entering Condition F (restore the inoperable channel within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ) when a train is inoperable instead of Condition C (restore the inoperable channel within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Differences Between the Proposed Change and the ISTS None Summary of the ISTS Justification The P-4 interlock is enabled when a reactor trip breaker (RTB) and its associated bypass breaker are open. Once the P-4 interlock is enabled, if a safety injection (SI) has occurred, reset of the SI is allowed after a 60 second time delay. This Function allows operators to take manual control of SI systems after the initial phase of injection is complete. Once the SI is reset, automatic actuation of SI cannot occur until the RTBs have been manually closed. The functions interlocked with P-4 avert or reduce the continued cooldown of the reactor coolant system (RCS) following a reactor trip. An excessive cooldown of the RCS following a reactor trip could cause an insertion of positive reactivity with a subsequent increase in generated power. Addition of feedwater to a steam generator associated with a steamline or feedline break could result in excessive containment building pressure. To avoid such a situation, the functions have been interlocked with P-4 as part of the design of the control and protection system.

If a P-4 Interlock channel is inoperable, a Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is proposed to return it to Operable status. The proposed Completion Time is reasonable considering the nature of the functions, the available redundancy, and the low probability of an event occurring during this interval. If the ESFAS P-4 Interlock cannot be returned to Operable status within the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time, the unit must be placed in Mode 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems.

During conversion to the ISTS, Farley chose to retain the existing Actions for an inoperable P-4 channel, as described in the response to RAI 3.3.2-5, changes to JFD 3 to specification 3.3.2. As described in the response to RAI 3.3.2-5, selection of Condition C as the applicable condition for Table 3.3.2-1, Function 7b, was based on the licensing basis defined in the Farley Custom Technical Specifications, current plant practices, and the applicable setpoint uncertainty calculations.

Differences Between the Plant-Specific Justification and the ISTS Justification None El - 94

Enclosure 1 Basis for Proposed Changes Licensee Commitments Required to Adopt this Change None NRC Approval The Westinghouse ISTS has provided the same Condition F actions and Completion Times for an inoperable P-4 channel since Revision 0 of the ISTS was issued in 1992 and since Revision 4 of NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors," was issued in 1981. Plant specific examples of the Condition F Actions and Completion Times are the Vogtle and North Anna Technical Specifications.

List of Affected Paqes 3.3.2-12 B3.3.2-33 B3.3.2-36 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met. The proposed change makes the remedial actions consistent with safety significance of the condition when a P-4 instrumentation channel is inoperable. The proposed changes are consistent with the ISTS for Westinghouse Plants (NUREG-1431).

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the Condition to be entered when the ESFAS Interlock P-4 is inoperable. Current Technical Specifications El - 95

Enclosure 1 Basis for Proposed Changes require restoring the channel to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in Mode 3 within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 5 within the following 52 hours6.018519e-4 days 0.0144 hours 8.597884e-5 weeks 1.9786e-5 months .

The proposed change provides 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to restore the inoperable channel, or be in Mode 3 in 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months and Mode 4 in 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months . The ESFAS P-4 interlock is not an initiator to any accident previously evaluated. The consequences of any accident previously evaluated during the proposed Completion Time are no different from the consequences during the existing Completion Time. As a result, the proposed change does not result in a significant increase in the consequences of any accident previously evaluated. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change provides an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore an inoperable ESFAS P-4 Interlock. During the proposed Completion Time, manual actions can perform the functions provided by the inoperable P-4 interlock. Also, the proposed Completion Time is reasonable given the available redundant channel, and the low probability of an event occurring during this interval. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

El - 96

Enclosure 1 Basis for Proposed Changes 2.24 Vogtle Consistency Change #1 - Revise LCO 3.5.5 to 8-hour Completion Time and Note allowance same as Vogtle Description of Proposed Change The proposed change modifies the LCO 3.5.5, "Seal Injection Flow," Action A, "Seal injection flow not within limit," Completion Time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and the Note to SR 3.5.5.1 to allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to stabilize reactor coolant system (RCS) pressure prior to verifying the seal injection throttle valves are properly adjusted.

Differences Between the Proposed Change and the ISTS ISTS 3.5.5, which is equivalent to Farley Specification 3.5.5, provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time when seal injection flow is not within limit and the SR Note allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to stabilize RCS pressure prior to performing the test.

Summary of the Approved Vogtle Justification The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for LCO 3.5.5, Required Action A.1 and the Note to SR 3.5.5.1 is revised to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 4-hour limit was based on the pre-ISTS "leakage" definition under which seal injection flow was considered operational leakage. Under that definition, the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time was consistent with the Specification 3.4.13, "RCS Operational Leakage," Completion Time for operational leakage not within limit. However, under the ISTS, seal injection flow is explicitly excluded from operational leakage. See the definition of "leakage," in Section 1.1 of the ISTS and the Farley Technical Specifications. Therefore, it is not necessary to maintain a Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for internal consistency and the Completion Time and surveillance Note time allowance should be based on maintaining plant safety and operational requirements.

The proposed 8-hour limit provides additional time to perform SR 3.5.5.1 and to make any necessary adjustments after RCS pressure has stabilized. The additional time is acceptable on the basis that there is little likelihood of an event that would challenge the ECCS occurring during the 8-hour window, and it reduces the pressure on the operations staff should iterations in the adjustment procedure be necessary to balance seal injection flow.

Differences Between the Plant-Specific Justification and the Vogtle Justification The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is conservative compared to other, comparable Completion Times in the Specifications. For example, Specification 3.5.2, "ECCS

- Operating," provides a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for one inoperable train of ECCS provided there is flow the equivalent to a single Operable train. Incorrectly adjusted seal injection flow would not be sufficient to reduce post-accident ECCS flow (assuming a single failure does not occur) to less than that of a single train of ECCS. Therefore, the proposed 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is more restrictive than the Completion Time in Specification 3.5.2 for a less safety significant condition.

The Completion Time and the SR 3.5.5.1 Note allowance should provide the same time period to determine seal injection flow and to perform any necessary adjustments.

El - 97

Enclosure 1 Basis for Proposed Changes Licensee Commitments Required to Adopt this Change None NRC Approval The NRC approved the proposed change for Vogtle in Amendment Numbers 96/74, dated September 25, 1996 (ACN 9610090381).

List of Affected Pages 3.5.5-1 3.5.5-2 B3.5.5-3 B3.5.5-4 Applicable Regulatory Requirements/Criteria Title 10 of the Code of Federal Regulations (10 CFR), 10 CFR 50.36(c)(2), states:

Limiting conditions for operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

10 CFR 50.36(c)(3), states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

There is no regulatory requirement that specifies what remedial actions are to be taken when an LCO is not met or what Completion Times are required. There is no regulatory requirement that specifies how Surveillance Requirements will be presented in the Technical Specifications. The proposed change makes the remedial actions and Surveillance Requirement consistent with safety significance of the condition when a P-4 instrumentation channel is inoperable.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

El - 98

Enclosure 1 Basis for Proposed Changes Response: No.

The proposed change modifies the LCO 3.5.5, "Seal Injection Flow,"

Action A, "Seal injection flow not within limit," Completion Time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and the Note to SR 3.5.5.1 to allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to stabilize reactor coolant system (RCS) pressure prior to verifying the seal injection throttle valves are properly adjusted. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility. Seal injection flow is not an initiator of any accident previously evaluated. The consequences of any accident previously evaluated during the extended Completion Time or Note allowance are the same as during the existing Completion Time and Note allowance. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change provides additional time to verify seal injection flow is within limit or to restore seal injection flow to within limit if it is discovered that it is not within limit. The additional time is acceptable on the basis that there is little likelihood of an event that would challenge the ECCS occurring during the 8-hour window, and it reduces the pressure on the operations staff should iterations in the adjustment procedure be necessary to balance seal injection flow. Therefore, this proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

El - 99

Enclosure 1 Basis for Proposed Changes 3.0 Environmental Considerations SNC has reviewed the proposed changes pursuant to 10 CFR 50.92 and determined that it does not involve a significant hazards consideration. In addition, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure. Consequently, the proposed Technical Specifications changes have no significant effect on the human environment and satisfy the criteria of 10 CFR 51.22 for categorical exclusion from the requirements for an environmental assessment.

El - 100

Joseph M. Farley Nuclear Plant - Units 1 and 2 Request for Technical Specification Amendment Adoption of Previously NRC-Approved Generic Technical Specification Changes and Other Changes Enclosure 2 Marked-Up Technical Specifications Pages

Enclosure 2 Marked-Up Technical Specifications Pages Index of Affected Technical Specification Pages vs. Traveler or Change Traveler(s) or Page TSTF-439-A 3.8.1-3 Change TSTF-283-A 3.8.1-8 TSTF-248-A 1.1-5 TSTF-283-A 3.8.1-9 TSTF-439-A 1.3-2 TSTF-283-A 3.8.1-12 TSTF-439-A 1.3-6 TSTF-283-A 3.8.1-13 TSTF-439-A 1.3-7 TSTF-439-A 3.8.9-1 TSTF-284-A 1.4-1 TSTF-272-A 3.9.1-1 TSTF-284-A 1.4-4 TSTF-312-A 3.9.3-1 TSTF-314-A 3.1.4-2 TSTF-284-A 3.9.3-2 TSTF-315-A 3.1.8-1 TSTF-349-A 3.9.5-1 TSTF-314-A 3.2.4-1 TSTF-308-A 5.5-3 TSTF-314-A 3.2.4-3 TSTF-273-A 5.5-13 TSTF-371-A 3.3.1-9 TSTF-343-A 5.5-14 TSTF-355-A 3.3.1-14 TSTF-355-A 3.3.1-15 TSTF-355-A 3.3.1-16 TSTF-355-A 3.3.1-17 TSTF-355-A 3.3.1-18 TSTF-355-A 3.3.1-19 TSTF-355-A 3.3.1-20 TSTF-355-A 3.3.1-21 TSTF-355-A 3.3.2-9 TSTF-355-A 3.3.2-10 TSTF-355-A 3.3.2-11 TSTF-355-A, 3.3.2-12 ISTS Adoption #1 TSTF-266-A 3.3.4-1 TSTF-266-A 3.3.4-3 TSTF-27-A 3.4.2-1 TSTF-87-A 3.4.5-2 TSTF-87-A 3.4.9-1 TSTF-247-A 3.4.11-1 TSTF-247-A, 3.4.11-3 TSTF-284-A TSTF-284-A 3.4.12-4 TSTF-325-A 3.5.2-1 Vogtle Change #1 3.5.5-1 Vogtle Change #1 3.5.5-2 TSTF-46-A 3.6.3-6 TSTF-439-A 3.6.6-1 TSTF-340-A, 3.7.5-1 TSTF-439-A TSTF-245-A 3.7.5-2 TSTF-245-A 3.7.5-3 TSTF-439-A 3.8.1-2 E2 - 1

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Definitions 1.1 1.1 Definitions PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. I QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2775 MWt.

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are T -248 fully inserted except for the single RCCA of highest However, with all RCCAs verified reactivity worth, which is assumed to be fully withdrawn.

fully inserted by two independent th any RCCA not capable of being fully inserted, the means, it is not necessary to reactivity worth of the RCCA must be accounted for in account for a stuck RCCA in the the determination of SDM; and SDM calculation.

(continued)

Farley Units 1 and 2 1.1-5 Amendment No. F (Unit 1)

Amendment No. P, .l(Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Completion Times 1.3 1.3 Completion Times DESCRIPTION limits, the Completion Time(s) may be extended. To apply this (continued) Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concurrent with the first inoperability; and

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Examiipii 1.3-3 illubtI dtr u0le uS u0 this typo of Cmpletio m The 10 day Completiorn Tiom specified for CGnJ*tpes A armd B in Example 1.3 3 .mabe met . exte.ded.

Farley Units 1 and 2 1.3-2 Amendment No. F (Unit 1)

Amendment No.J (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1-3-3 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore 7 days Function X Function X train train to OPERABLE inoperable, status.

10 days from discovery of failul e t meet1 he I*G4 B. One B.1 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function Y Function Y train train to OPERABLE inoperable, status.

C. One C.1 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function X Function X train train to OPERABLE inoperable, status.

AND OR One C.2 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function Y Function Y train train to OPERABLE inoperable, status.

(continued)

Farley Units I and 2 1.3-6 Amendment No. r (Unit 1)

Amendment No. Iý (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A. 1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A). IT The C~ompletiom Timeos ef Coirlditiena A and B ore medified by a logieal DOnnectGr with a separat9 10 day Completion Tim-e mea2&sured from the

..... was.t dise.ver.d the LCO w not

-a et. In this xample, withut th separate Completion Time, it weuld be pooaiblc to altefrnate betwecen Gemdfitios A, B, and C in atuch a mianner that epelratien could continue 4mdefinitely without eveF FezteoRng cyatoeFA to lmoot the_ '_G_. The separate ComRplotion Time modified by tho phrase fro m.dise~eo~o of falr oiee h~ei desiid to preveirt -ndefinite eeintimued epefetien while not meeting the LCO. This Completion Time allows for an exception to thei normal2 "tiezeoG" for beginning the ComRpletion TimeG

..ULpk". 1, t fis ins~tal ic, the C 1 petio T~ip ii "tome zero" is speeified as Gemmenreing at the tiame the LCO was initially not met, instead of at the tame the asseoiatod Condition was eantered.

Insert -TS1.

Examples (continued)

Farley Units 1 and 2 1.3-7 Amendment No. G (Unit 1)

Amendment No. ffq (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Insert - TS 1.3 Example It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO.

However, doing so would be inconsistent with the basis of the Completion Times.

Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the lInsert - TS 1.4 Description Surveillance column that modify performance requirements. ITST F-284 1 n1-1 Situatiens wherc a Sur.'zillaneo coeuld bo roquirod (i.e., its FrFequency acould- expire), but whera it is not possible or not d-esire-d that it be perfermed utinfl semetime aftelr the asseoiated LCO is withi its Applicability, leprecont potential SR 3.0.4 cGIGictc. To avoid thore eenfliets, the SR (i.e., the Survcillanoz 8Frthe FrFequency) us -tatod rucwh that it Ison!'; "rcguircd" when it can be and should be por9Formod. With an1 SR Satisfied, SR~ 3.0.4 imrpeses fie rzctritieon.

EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

(continued)

Farley Units 1 and 2 1.4-1 Amendment No. ER (Unit 1)

Amendment No. R- (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages INSERT- TS 1.4 Description ITSTF-284 Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied:

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or

b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or

c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.

Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLEF 1-4-3 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY NOTE ------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

> 25% RTP.

Perform channel adjustment. 7 days The interval continues, whether or not the unit operation is < 25% RTP between performances.

As the Note modifies the required per[ormance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches > 25% RTP to perform the Surveillance.

The Surveillance is still considered to be performed within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was

< 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with power > 25% RTP.

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for Insert - TS 1.4 completing the Surveillance. If the Surveillance were not performed Example 1.4-4 within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval, there would then be a failure to perform a SIlnsert - TS 1.4 ; Surveillance within the specified Frequency, and the provisions of ~284 Example 1.4-5 SR 3.0.3 would apply.

Insert - TS1.4-6 lExample 1.4-6 /

Farley Units 1 and 2 1.4-4 Amendment No. [i (Unit 1)

Amendment No. F-J (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Insert - TS 1.4 Example 1.4-4 ITTF284I EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Only required to be met in MODE 1.

Verify leakage rates are within limits. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.

to NL-14-1385 Marked-Up Technical Specifications Pages Insert - TS 1.4 Example 1.4-5 ITTF284I EXAMPLE 1.4-5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

.-------------------. NOTE--------------

Only required to be performed in MODE 1.

Perform complete cycle of the valve. 7 days The interval continues, whether or not the unit operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances.

As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1.

Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.

Once the unit reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed. If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

to NL-14-1385 Marked-Up Technical Specifications Pages Insert - TS 1.4 Example 1.4-6 JTSTF284 EXAMPLE 1.4-6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Not required to be met in MODE 3.

Verify parameter is within limits. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the unit is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3). The interval measurement for the Frequency of this Surveillance continues at all times. As described in Example 1.4-1. However. the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore. if the Surveillance were not performed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval (plus the extension allowed by SR 3.0.2). and the unit was in MODE 3, there would be no failure of the SR nor failure to meet the LCO. Therefore. no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION R REQUIRED ACTION I COMPLETION TIME B. (continued) B.2.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND 8.2.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to < 75% RTP.

AND B.2.3 Verify SDM to be within Once per the limits provided in,the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months COLR.

AND Po Rand SR 3.2.1.2 B.2.4 Perform SR 3.2.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not met.

Farley Units 1 and 2 3.1.4-2 Amendment No. R (Unit 1)

Amendment No. - (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages PHYSICS TESTS Exceptions-MODE 2 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions-MODE 2 LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of LCO 3.1.3, "Moderator Temperature Coefficient (MTC)"; TSTF-315 I LCO 3.1.4, "Rod Group Alignment Limits"; and t he number of LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO red channels for 3.1.6, "Control Bank Insertion Limits"; and requil LCO 3.4.2, "RCS Minimum Temperature for Criticality" 3.3.1, "RTS InstrL imentation,"

may be suspended, rovided: Func1tions 2, 3, and 17.e, may be reduced

a. THERMAL POWER is <s5% RTP;It3

b. SDM is within the limits provided in the COLR; and

c. RCS lowest loop average temperature is a 531°F.

APPLICABILITY: MODE 2 during PHYSICS TESTS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to Immediately restore SDM to within limit.

AND A.2 Suspend PHYSICS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TESTS exceptions.

B. THERMAL POWER not B.1 Open reactor trip Immediately within limit, breakers.

Farley Units 1 and 2 3.1.8-1 Amendment No. F-'(Unit 1)

Amendment No. [J(Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages QPTR 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be < 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. QPTR not within limit. A.1 Limit THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each POWER to > 3% below QPTR determination RTP for each 1% of QPTR > 1.00.

AND A.2 Determine QPTR. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND FTsTF- 314 I A.3 Perform SR 3.2.1.1 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after SR 3.2.2.1. achieving equilibrium conditions with SR 3.2.1.2, THERMAL POWER limited by Required Action A. 1 AND Once per 7 days thereafter ANDi (continued)

Farley Units 1 and 2 3.2.4-1 Amendment No. rit (Unit 1)

Amendment No. F (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages QPTR 3.2.4 ACTIONS CONDITION I REQUIRED ACTION COMPLETION TIME A. (continued) A.6 -- ----------

NOTE------

Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ITSF314 SR 3.2.2.1. achieving equilibrium conditions at RTP

SR 3.2.1.2, OR Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of Required Action A. 1 B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 50% RTP.

Time not met.

Farley Units 1 and 2 3.2.4-3 Amendment No. 5 (Unit 1)

Amendment No. F (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 ---------- --- NOTE ----------------

Not required to be performed for source range instrumentation until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < P-6.

Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 ---- -------------- NOTES-

i. A dju tLNis .... .... Of'.... . ..... ca .. ....... "

power range channel pow Ol NISIXced16 i poWer by moroU IdicItId output. Adjust power then 4-2% RTP.L range channel output if calorimetric heat Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months balance calculation after THERMAL POWER is >_15% RTP.

results exceed power In accordance with range channel output the Surveillance by more than +2%.

Frequency Control Program TSTF-371 SR 3.3.1.3 --------------------------- NOTES---------------

A ,djui4t l hnltll nzl. if.a lutc dfll rznt zi, 2THERMAL M. Not required to be performed until 7 days after POWER is > 50% RTP.

Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the incore detector In accordance with measurements t AFD., iAdjust NIS channel if the Surveillance Frequency Control INuclear Instrumentation System (NIS)J difference is >/= 3%. Program I______________

Farley Units 1 and 2 3.3.1-9 Amendment No. IlJ (Unit 1)

Amendment No. FI (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE ' TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual Reactor 1,2 2 B SR 3.3.1.12 NA NA Trip 3 (a), 4 (a), 5 (a) 2 C SR 3.3.1.12 NA NA

2. Power Range Neutron Flux

a. High 1,2 D SR 3.3.1.1

109.4% RTP [109%

SR 3.3.1.2 RTP SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.14

b. Low E SR 3.3.1.1 25.4% RTP [E25% RTP SR 3.3.1.8 SR 3.3.1.10 SR 3.3.1.14

3. Power Range 1,2 4 D SR 3.3.1.7 _ 5.4% RTP -R5%RTP Neutron Flux High SR 3.3.1.10 with time with time Positive Rate SR 3.3.1.14 constant constant

_ 2 sec > 2 sec

4. Intermediate 1(b), 2 (c) 2 F,G SR 3.3.1.1 _ 40% RTP R-35% RTP Range Neutron SR 3.3.1.8 Flux SR 3.3.1.10 2 (d) 2 H SR 3.3.1.1 < 40% RTP E935% RTP SR 3.3.1.8 SR 3.3.1.10 (a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b) Below the P-10 (Power Range Neutron Flux) interlocks.

(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

Farley Units 1 and 2 3.3.1-14 Amendment No. * (Unit 1)

Amendment No. * (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE ' TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

5. Source Range (d) 2 2 lI,J SR 3.3.1.1 < 1.3 E5 cps EP.0 E5 cps Neutron Flux SR 3.3.1.8 SR 3.3.1.10 3 (a), 4 (a), 5 (a) 2 J,K SR 3.311.1

1.3 E5 cps IE1.0 E5 cps SR 3.3.1.7 SR 3.3.1.10 3 (e) 4 (e), 5 (e) L SR 3.311.1 1 N/A N/A SR 3.3.1.10

6. Overtemperature 1,2 3 E SR 3.311.1 Refer to Refer to AT SR 3.3.1.3 Note 1 (Page Note 1 (Page SR 3.3.1.7 3.3.1-20) 3.3.1-20)

SR 3.3.1.9 SR 3.3.1.10 SR 3.3,1.14

7. Overpower AT 1,2 3 E SR 3.3.1.1 Refer to Refer to SR 3.3.1.7 Note 2 (Page Note 2 (Page SR 3.3,1.10 3.3.1-21) 3.3.1-21)

SR 3.3.1.14 (a) With RTBs closed and Rod Control System capable of rod withdrawal.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does provide indication.

Farley Units 1 and 2 3.3.1-15 Amendment No. 5* (Unit 1)

Amendment No. r (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

8. Pressurizer Pressure

a. Low 1 (t) 3 M SR 3.3.1.1 > 1862 psig 1 865 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.14

b. High 1,2 3 E SR 3.3.1.1 5 2388 psig [E2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.14 M f

9. Pressurizer Water 3 M SR 3.3.1.1 592.4% 1j92%

Level - High SR 3.3.1.7 SR 3.3.1.10

10. Reactor Coolant 1(f) 3 per loop M SR 3.3.1.1 > 89.7% G 90%

Flow- Low SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.14 (f) Above the P-7 (Low Power Reactor Trips Block) interlock.

Farley Units 1 and 2 3.3.1-16 Amendment No. 14 (Unit 1)

Amendment No. (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

11. Notused 11(f) > 2640 V B 2680 V

12. Undervoltage 3 M SR 3.3.1.6 RCPs SR 3.3.1.10 1(f)

13. Underfrequency 3 M SR 3.3.1.6 >_56.9 Hz r"-57 Hz RCPs SR 3.3.1.10

14. Steam 1,2 3 per SG E SR 3.3.1.1 _ 27.6% g28%

Generator (SG) SR 3.3.1.7 Water Level - SR 3.3.1.10 Low Low SR 3,3.1.14 (1) Above the P-7 (Low Power Reactor Trips Block) interlock.

Farley Units 1 and 2 3.3.1-17 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

15. Turbine Trip

a. Low Auto Stop 1(i) 3 P SR 3.3.1.10 ? 43 psig EJ45 psig I Oil Pressure SR 3.3.1.13

b. Turbine Throttle 1 (i) 4 Q SR 3.3.1.10 NA NA I Valve Closure SR 3.3.1.13 1,2

16. Safety Injection (SI) 2 trains R SR 3.3.1.12 NA NA I Input from Engineered Safety Feature Actuation System (ESFAS)

17. Reactor Trip System Interlocks

a. Intermediate 2 (d) 2 T SR 3.3.1.10 _6E-11 amp rBlE-10 Range Neutron SR 3.3.1.11 amp Flux, P-6

b. Low Power 1 1 per train U NA NA NA I Reactor Trips Block, P-7

c. Power Range 1 4 U SR 3.3.1.10 < 30.4% RTP [330%RTP I Neutron Flux, SR 3.3.1.11 P-8

d. Power Range 1 4 U SR 3.3.1.10 < 50.4% RTP B50%RTP I Neutron Flux, SR 3.3.1.11 P-9

e. Power Range 1,2 4 T SR 3.3.1.10 Ž7.6% RTP a8% RTP Neutron Flux, SR 3.3.1.11 and and P-10 -< 10.4% RTP [D 0% RTP

f. Turbine Impulse 2 U SR 3.3.1.1 _511% E]1O% I Pressure, P-13 SR 3.3.1.10 turbine turbine SR 3.3.1.11 power power (d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(i) Above the P-9 (Power Range Neutron Flux) interlock.

Farley Units 1 and 2 3.3.1-18 Amendment No.r (Unit 1)

Amendment No. r (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

18. Reactor Trip 1,2 2 trains S, W SR 3.3.1,4 NA NA Breakers (J) 3 (a), 4 (a), 5 (a) 2 trains C, W SR 3.3.1.4 NA NA 19, Reactor Trip 1,2 1 each per V SR 3.3.1.4 NA NA Breaker RTB Undervoltage and 3 (a), 4 (a). 5(a) C SR 3.3.1.4 NA NA Shunt Trip 1 each per Mechanisms RTB

20. Automatic Trip 1,2 2 trains R, W SR 3.3.1.5 NA NA I Logic 3 (a), 4 (a), 5 (a) 2 trains C, W SR 3.3.1.5 NA NA I (a) With RTBs closed and Rod Control System capable of rod withdrawal.

j) Including any reactor trip bypass breaker that is racked in and closed for bypassing an RTB.

Farley Units I and 2 3.3.1-19 Amendment No. rG--l (Unit 1)

Amendment No. 11 (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 8)

Reactor Trip System Instrumentation Note 1: Overtemperature AT [nominal The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 0.4% of AT span.

AT 0 + TOs) '5ATf lK (I+ r 1s) [

0 + r 5 s)Td{I2(i +rs)S (I + TOs)

I -"T'I+K3(P-jP')-fj(AI)t Where: AT is measured loop AT, OF.

AT 0 is the indicated loop AT at RTP and reference Tavg, OF.

s is the Laplace transform operator, sec-1.

T is the measured loop average temperature, OF.

T' is the reference Tavg at RTP, <

OF.

P is the measured pressurizer pressure, psig.

P' is the nominal pressurizer operating pressure =

psig.

K1, K2 = */°F K3 **/psi T' >

sec T2 <

sec 1=

sec T 5 <

sec T6 -

sec f1 (Al) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

fl(AI) = *{*+ (qt - qb)} when (qt - qb)<- * % RTP

% of RTP when *% RTP < (qt - qb)-* *% RTP

((q, - qb) -

when (qt- qb)> *% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

as specified in the COLR Farley Units 1 and 2 3.3.1-20 Amendment No. F; (Unit 1)

Amendment No. I--- (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 8)

Reactor Trip System Instrumentation Note 2: Overpower AT nominal The Overpower AT Function Allowable Value shall not exceed the followin Trip Setpoint by more than 0.4% of AT span.

A T QI+ r 4 S) < A To____T3 ) -K6 + TOs TI----I I f2 (A 1)}

(I + 0 1 +~sZ3S +4KTO Where: AT is measured loop AT, OF.

ATo is the indicated loop AT at RTP and reference Tavg, OF.

s is the Laplace transform operator, sec-1 .

T is the measured loop average temperature, OF.

T" is the reference Tavg at RTP, <

OF.

4

K5 = */OF for increasing Tavg K6 = */OF when T > T" K5 = */OF for decreasing Tavg K6 = */OF when T _<T" T3 _

sec

=

sec T5 <

sec T6 _

sec f 2(AI) = *% RTP for all Al.

as specified in the COLR Farley Units I and 2 3.3.1-21 Amendment No. [T (Unit 1)

Amendment No. (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Safety Injection

a. Manual Initiation 1,2,3,4 2 B SR 3.3.2.6 NA NA

b. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.8 Relays

c. Containment 1,2,3 3 D SR 3.3.2.1 _<4.5 psig [4.0 psig Pressure - SR 3.3.2.4 High 1 SR 3.3.2.7 SR 3.3.2.9 1,2,3(a)

d. Pressurizer 3 D SR 3.3.2.1 2t 1847 psig r1850 psig Pressure - Low SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9

e. Steam Line Pressure (1) Low 1,2,3(b) 1 per steam D SR 3.3.2.1 ,, 5 7 5 (c) psig -' 585 (c) psig line SR 3.3.2.4 SR 3.3.2.7 SR 3.3,2.9 (2) High 1,2,3 3 per steam D SR 3.3.2.1 !*l12 psig E9100Psig Differential line SR 3.3.2.4 Pressure SR 3.3.2.7 Between SR 3.3.2.9 Steam Lines (a) Above the P-1 1 (Pressurizer Pressure) interlock.

(b) Above the P-12 (Tavg - Low Low) interlock.

(c) Time constants used in the lead/lag controller are t, > 50 seconds and t2 5 5 seconds.

Farley Units 1 and 2 3.3.2-9 Amendment No. l (Unit 1)

Amendment No. r (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT VALUE SETPOINT

2. Containment Spray

a. Manual Initiation 1,2,3,4 2 B SR 3.3.2.6 NA NA

b. Automatic Actuation 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.8

c. Containment 1,2,3 4 E SR 3.3.2.1 _<28.3 psig B_27 psig Pressure SR 3.3.2.4 High - 3 SR 3.3.2.7 SR 3.3.2.9

3. Containment Isolation a, Phase A Isolation (1) Manual 1,2,3,4 2 B SR 3.3.2.6 NA NA Initiation (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.8 Relays (3) Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

b. Phase B Isolation (1) Manual 1,2,3,4 2 B SR 3.3.2.6 NA NA Initiation (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.8 Relays (3) Containment 1,2,3 4 E SR 3.3.2.1 !5 28.3 psig r27 psig Pressure SR 3.3.2.4 High - 3 SR 3.3.2.7 SR 3.3.2.9 Farley Units 1 and 2 3.3.2-10 Amendment No.i* (Unit 1)

Amendment No. I (Unit 2)I

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 4)

Engineered Safety Feature Actuation System Instrumentation ITSTF-355 APPLICABLE INominal MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Steam Line Isolation 1 per steam

a. Manual Initiation 1 ,2 (d) 3 (d) line F SR 3.3.2.6 NA NA

b. Automatic 1 ,2 (d),3 (d) 2 trains G SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2,3 and Actuation SR 3.3.2,8 Relays

c. Containment 1 ,2 (d), 3 (d) 3 D SR 3.3.2.1 < 17.5 psig [E16.2 psig Pressure - High 2 SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9

d. Steam Line 1,2 (d), 3 (b)(d) 1 per steam D SR 3.3.2.1 > 575(c) psig [!1 5 8 5 (c) psig Pressure Low line SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9

e. High Steam Flow 1 ,2 (d), 3 (d) 2 per steam D SR 3.3.2.1 (e) (f) in Two Steam line SR 3.3.2.4 Lines SR 3.3.2.7 Coincident with 1 ,2 (d), 3 (d) 1 per loop D SR 3.3.2.1 >542.6°F ED543oF Tavg - Low Low SR 3.3.2.4 SR 3.3.2.7 (b) Above the P-12 (Tavg - Low Low) interlock.

(c) Time constants used in the lead/lag controller are tj > 50 seconds and t2 < 5 seconds.

(d) Except when one MSIV is closed in each steam line.

(e) Less than or equal to a function defined as AP corresponding to 40.3% full steam flow below 20% load, AP increasing linearly from 40.3% full steam flow at 20% load to 110.3% full steam flow at 100% load.

(f) Less than or equal to a function defined as AP corresponding to 40% full steam flow between 0% and 20% load and then a AP increasing linearly from 40% steam flow at 20% load to 110% full steam flow at 100% load.

Farley Units 1 and 2 3.3.2-11 Amendment No. i (Unit 1)

Amendment No. * (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE inal MODES OR Nominal OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

5. Turbine Trip and Feedwater Isolation 1,2 2 trains H SR 3.3.2.2 NA NA

a. Automatic Actuation Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.8

b. SG Water Level - 1,2 3 per SG SR 3.3.2.1 582.4% B82%

High High (P-14) SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

6. Auxiliary Feedwater 1,2,3 2 trains G SR 3.3.2.2 NA NA

a. Automatic Actuation SR 3.3.2.3 Logic and Actuation SR 3.3.2.8 Relays

b. SG Water Level - 1,2,3 3 per SG D SR 3.3.2.1 > 27.6% D 28%

Low Low SR 3.3.2.4 SR 3.3.2.7 SR 3 .3.2.9(g)

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

d. Undervoltage 1,2 3 I SR 3.3.2.5 2 2640 volts E92680 Reactor Coolant SR 3.3.2.7 volts Pump SR 3.3.2.9

e. Trip of all Main 1 2 per pump J SR 3.3.2.10 NA NA Feedwater Pumps

7. ESFAS Interlocks 1,2,3 2 trains L SR 3.3.2.2 NA NA

a. Automatic Actuation Logic and Actuation SR 3.3.2.3 b.

Relays Reactor Trip, P-4 1,2,3 1 per train, 2 F SR 3.3.2.8 SR 3.3.2.6 NA NA iSTS Adoption trains

c. Pressurizer 1,2,3 3 K SR 33.2.4 *2003 psig (3 2000 psig Pressure, P-1I SR 3.3.2.7

d. Tavg - Low Low, P-12 1,2,3 1 per loop K SR 3.3.2.4 2!542.6*F ID543*F SR 3.3.2.7 !* 545.4*F (3545-F (Decreasing)

(Increasing)

(g) Applicable to MDAFW pumps only.

Farley Units 1 and 2 3.3.2-12 Amendment No. = (Unit 1)

Amendment No. f (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Remote Shutdown System 3.3.4 3.3 INSTRUMENTATION 3.3.4 Remote Shutdown System LCO 3.3.4 The Remote Shutdown System Functionsli, fTbk: 85,3.4- ii shall be FTSTF-266I OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTE -----

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable, to OPERABLE status.

B. -------- NOTE -------- B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable to Source Range Neutron Flux AND function.

B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action and associated Completion Time not met.

Farley Units I and 2 3.3.4-1 Amendment No. 2 (Unit 1)

Amendment No. F (Unit 2) to NL-14-1385 Marked-Up Te Removed from TS and placed in Bases A ITSTF266 Remote Shutdown System as Table B 3.3.4-1 3.3.4 Table 3.3.4-1 (page 1 of 1)

Remote Shutdown System Instrumentation and Controls

\,,FUNCTION/INSTRUMENT REQUIRED lRCONTROL PARAMETER NUMBER OF CHA/ELS

1. Steam Generator ide Range Level 1/SG

2. Steam Generator Pre ure 1/SG

3. Pressurizer Water Level 1

4. Pressurizer Pressure 1

5. RCS Hot Leg Temperature (Lo A) 1

6. RCS Cold Leg Temperature (Loo 1

7. Source Range Neutron Flux (Gammi etrics) 1

8. Condensate Storage Tank Level 1

9. Reactivity Control

a. Boric Acid Transfer stem 1

10. RCS Pressure

a. Pressurizer H er Control 1

11. RCS Inventory

a. Chargin ystem

b. Letdo n Orifice Isolation Valves 1

12. Decay H t Removal

a. uxiliary Feedwater System

, SG Atmospheric Relief Valves

13. afety Grade Support Systems Required For Functions Listed Above Farley Units 1 and 2 3.3.4-3 Amendment No. ýJ (Unit 1)

Amendment No. F (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.2 Each RCS loop average temperature (Tavg) shall be > 541 OF.

APPLICABILITY: MODE 1, MODE 2 with kffŽ- 1.0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Tavg in one or more RCS A.1 Be in MODE 3. 30 minutes loops not within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each loop > 541 OF. NOTE In accordance with low e v, .......

th e S u rve illa n ce R. - . ,. , . _ n... .

Frequency Control Res , , avg Program 30im*,fes the~eafteF Farley Units 1 and 2 3.4.2-1 Amendment No. (Unit 1)

Amendment No. IM3 (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages RCS Loops -MODE 3 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One required RCS loop C.1 Restore required RCS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not in operation, loop to operation.

Place the Rod Control System in a el¢sed and Rod Control OR ,---condition incapable of rod withdrawal.

System callble of rod withdrawal. with C.2 .D crgiuz

. .. ll con.tro 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months rod drive machanWmsIT fGRaMe) - ----- FT- 87I D. Two required RCS loops D. Two required RCS loops D.1 I~~~~-1 tnrgzcal RDAc Immediately inoperable. Place the Rod Control System in a AND condition incapable of rod withdrawal.

OR D.2 Suspend all operations Immediately No RCS loop in involving a reduction of operation. RCS boron concentration.

AND D.3 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify steam generator secondary side water levels In accordance with are > 30% (narrow range) for required RCS loops, the Surveillance Frequency Control Program SR 3.4.5.3 Verify correct breaker alignment and indicated power In accordance with are available to the required pump that is not in the Surveillance operation. Frequency Control Program Farley Units 1 and 2 3.4.5-2 Amendment No. (Unit 1)

Amendment No. Ft8J (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level !563.5% indicated; and

b. Two groups of pressurizer heaters OPERABLE with the capacity of each group ->125 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

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

Pressurizer water level limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level A.1 Be in MODE 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months IF:r*F trip reý2=

not within limit.

F~T__87__

INSERT - TS 3.4.9 Condition A AND

'AO/*Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. One required group of B.1 Restore required group of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters pressurizer heaters to inoperable. OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not AND met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Farley Units 1 and 2 3.4.9-1 Amendment No. j (Unit 1)

Amendment No. l (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages INSERT - TS 3.4.9 Condition A FTsTF-8-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.2 Fully insert all rods. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.3 Place Rod Control 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months System in a condition incapable of rod withdrawal.

AND to NL-14-1385 Marked-Up Technical Specifications Pages Pressurizer PORVs 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS ITTF247I I.I IrI

I',ILJI r -------- ----- - ---- - ---- - ----------

Separate Condition entry is allowed for each PORV.

andeach block valveý CONDITION REQUIRED ACTION COMPLETION TIME A. One or more PORVs A.1 Close and maintain 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable and capable of power to associated being manually cycled, block valve.

B. One PORV inoperable and B. 1 Close associated block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not capable of being valve.

manually cycled.

AND B.2 Remove power from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated block valve.

AND B.3 Restore PORV to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

Farley Units 1 and 2 3.4.11-1 Amendment No. I (Unit 1)

Amendment No. * (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Pressurizer PORVs 3.4.11 ACTIONS CONDITIONNi, 1I REQUIRED ACTION COMPLETION TIME F. I;.j.o. oncock F.'I ,-,, e 999e6i:,z 1--he9 -

inoperable. roRV9 in moenual valves ITTF247 I

_1 Restore one block valve 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Ui1 to OPERABLE status.

bleek valve-to E)PERABLE status.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition F not AND met.

G.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Iperformed I I

I----*.*

NOTES - ---------- - I SR 3.4.11.1

1. Not required to be"i ewith block valve closed in accordance with the Requireft A -

Indien 13 ' Actions of this 2.1 Not requied to be pe ri med pi tt entry ,,to I JTSTF-28i4 I IME)BE-8 I A

Perform a complete cycle of each block valve. In accordance with the Surveillance Only required to be performed in MODES Frequency Control 1 and 2. Program Farley Units 1 and 2 3.4.11-3 Amendment No. i (Unit 1)

Amendment No. Pr (Unit 2)

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of one charging pump is In accordance with capable of injecting into the RCS when one or more the Surveillance RCS cold legs is < 180°F. Frequency Control Program SR 3.4.12.2 Verify a maximum of two charging pumps are In accordance with capable of injecting into the RCS when all RCS cold the Surveillance legs are > 180 0 F. Frequency Control Program SR 3.4.12.3 Verify each accumulator is isolated. In accordance with the Surveillance Frequency Control Program SR 3.4.12.4 Verify RHR suction isolation valves are open for each In accordance with required RHR suction relief valve, the Surveillance Frequency Control Program I

SR 3.4.12.5 ------------------- NOTE ----------------

Only required to be ,cmplying

-- with ITS% 4 LCO 3.4.12.b. met Verify RCS vent > 2.85 square inches open. In accordance with the Surveillance Frequency Control Program SR 3.4.12.6 Verify each required RHR suction relief valve In accordance with setpoint. the Inservice Testing Program AND In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.12-4 Amendment No. r--(Unit 1)

Amendment No. f'](Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

t. IFrrr --------------------------------------------

IVJ.~ 'I*U

1. In MODE 3, the Residual Heat Removal or the Centrifugal Charging Pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2. Upon entry into MODE 3 from MODE 4, the breaker or disconnect device to the valve operators for MOVs 8706A and 8706B may be locked open for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to allow for repositioning from MODE 4 requirements.

I APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

AN&

At k0 st 1005b of thO EC CS_ flow oquiyalnt te a cinglo OPERABL6E EGGS trisin available.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

\ Insert - LCO 3,52 Condition C I Farley Units 1 and 2 3.5.2-1 Amendment No. 5; (Unit 1)

Amendment No. F (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages TS 3.5.2 - Condition C Insert CONDITION REQUIRED ACTION COMPLETION TIME C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

to NL-14-1385 Marked-Up Technical Specifications Pages Seal Injection Flow 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow LCO 3.5.5 Reactor coolant pump seal injection flow shall be within limits.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I 1ý:4 A. Seal injection flow not A. 1 Adjust manual seal Vogtle

Tours Change #1 within limit. injection throttle valves in accordance with SR 3.5.5.1.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Farley Units 1 and 2 3.5.5-1 Amendment No. F (Unit 1)

Amendment No. W (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE

FREQUENCY SR 3.5.5.1 IVogtle

-NOTE Not required to be performed unti Lhours after the Change #1 Reactor Coolant System pressure stabilizes at

> 2215 psig and < 2255 psig.

Verify manual seal injection throttle valves are In accordance with adjusted to give a flow within the limits of Figure the Surveillance 3.5.5-1 with the seal water injection flow control Frequency Control valve full open. Program Farley Units 1 and 2 3.5.5-2 Amendment No.Fr- (Unit 1)

Amendment No. * (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.3 - - --------.-...--

- - ------ N O TE S ---------- ------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. The blind flange on the fuel transfer canal flange is only required to be verified closed after each draining of the canal.

Verify each containment isolation manual valve and Prior to entering blind flange that is located inside containment and MODE 4 from not locked, sealed, or otherwise secured and MODE 5 if not required to be closed during accident conditions is performed within closed, except for containment isolation valves that the previous are open under administrative controls. 92 days

/---Power operated I I SR 3.6.3.4 Verify the4 olation time of eac power ý p;,; c; s, In accordance with automatic containment isolation valve in the IST the Inservice FTsTF-_4_67j Program is within limits. Testing Program SR 3.6.3.5 Perform leakage rate testing for containment In accordance with penetrations containing containment purge valves the Surveillance with resilient seals. Frequency Control Program AND Within 92 days after opening the valve SR 3.6.3.6 Verify each automatic containment isolation valve In accordance with that is not locked, sealed or otherwise secured in the Surveillance position, actuates to the isolation position on an Frequency Control actual or simulated actuation signal. Program Farley Units I and 2 3.6.3-6 Amendment No. r (Unit 1)

Amendment No. K=(Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Containment Spray and Cooling Systems 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray and Cooling Systems LCO 3.6.6 Two containment spray trains and two containment cooling trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A.1 Restore containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable, spray train to OPERABLE status. ____

10 dys froM di.,. erly of fIIlmIe to meetthe I GEe B. Required Action and B. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months C. One containment cooling C.1 Restore containment 7 days train inoperable, cooling train to OPERABLE status. ___

16 days homf discovery of fail,,e to meeee-*he-ee Farley Units 1 and 2 3.6.6-1 Amendment No. (Unit 1)

Amendment No. * (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

-NOTE.

LCO 3.0.4b is not applicable.

CONDITION A. One steam supply to turbine driven AFW ,[TSTF-340 pump inoperable.

B. One AFW train B. 1 Restore AFW train to inoperable for reasons I OPERABLE status.

other than Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A or B AND not met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR Two AFW trains inoperable.

Farley Units 1 and 2 3.7.5-1 Amendment No.rM (Unit 1)

Amendment No.r0 (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages Insert - TS 3.7.5 ITSTF-340-OR

NOTE------

Only applicable if MODE 2 has not been entered following refueling.

One turbine driven AFW pump inoperable in MODE 3 following refueling.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages AFW System 3.7.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Three AFW trains D.1 ------- NOTE-------

inoperable. LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately one AFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -NOTE INSERT - SR 3.7.5.1 ocn-"t"^l valves whe. +/-1--G% RT" e" v... th . AF .

Note s,2tem iE not in _u-mat"i w con..trol. ~245 Verify each AFW manual, power operated, and In accordance automatic valve in each water flow path, and in both with the steam supply flow paths to the steam turbine driven Surveillance pump, that is not locked, sealed, or otherwise Frequency secured in position, is in the correct position. Control Program Farley Units 1 and 2 3.7.5-2 Amendment No.r--R (Unit 1)

Amendment No. PJ (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages INSERT - SR 3.7.5.1 Note AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.2 NOTE -----------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after >_1005 psig in the steam generator.

Verify the developed head of each AFW pump at the In accordance flow test point is greater than or equal to the required with the Inservice developed head. Testing Program.

SR 3.7.5.3 /Verify each AFW automatic valve that is not locked, In accordance INSERT - SR 3753 sealed, or otherwise secured in position, actuates to with the lNote the correct position on an actual or simulated Surveillance actuation signal. Frequency Control Program SR 3.7.5.4 ------------------- NOTE ----------------- ITTF245

[w*NAot required to be performed for the turbine driven

- AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after > 1005 psig in the INSERT - SR 3.7.5.4 *_Tsteam generator.

Note Verify each AFW pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program SR 3.7.5.5 Verify the turbine driven AFW pump steam admission In accordance valves open when air is supplied from their respective with the air accumulators. Surveillance Frequency Control Program Farley Units 1 and 2 3.7.5-3 Amendment No.l'* (Unit 1)

Amendment No. FW (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages INSERT - SR 3.7.5.3 Note NOTE -------------

AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being ITSF245 I manually realigned to the AFW mode of operation.

INSERT - SR 3.7.5.4 Note ITTF2451

2. AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Enclosure 2 to NL-14-1385 Marked-Up Technical Specifications Pages AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months circuit to OPERABLE status. AND iieeoecy of filuro to-B. One DG set inoperable. ---------- NOTE---

LCO 3.0.4c is applicable when only one of the three DGs is inoperable.

B.1 Perform SR 3.8.1.1 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the required offsite circuit(s). AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s) supported by discovery of the inoperable DG set Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable, redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG set is not inoperable due to common cause failure.

OR (continued)

Farley Units 1 and 2 3.8.1-2 Amendment No.E (Unit 1)

Amendment No. F (Unit 2) to NL-14-1385 Marked-Up Technical Specifications Pages AC Sources - Operating 3.8.1 Af,,TIONR CONDITION REQUIRED ACTION JCOMPLETION TIME B. (continued) B.3.2 Perform SR 3.8.1.6 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG set.

AND ITSTF-439 I B.4 Restore DG set to 10 days OPERABLE status.

C. Two required offsite C.1 Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable. feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND C.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

Farley Units 1 and 2 3.8.1-3 Amendment No. FI (Unit 1)

Amendment No. f (Unit 2)

Enclosure 2 to NL-14-1 385 Marked-Up Technical However, this surveillance may be performed to reestablish OPERABILITY provided an AC Sources - Operating assessment determines the safety of the plant 3.8.1 is maintained or enhanced.

SURVEILLANCE REQUIREMENTS _ _ __

SURVEILL CE FREQUENCY SR 3.8.1.7 ---------------- ---NOTE ---------------------- ITSTF-283 I This Surveillanc shall note performed in MODE 1 or 2. Lnormally Verify manual transfer of AC power sources from the In accordance with normal offsite circuit to the alternate required offsite the Surveillance circuit. Frequency Control Program SR 3.8.1.8 Verify each DG rejects a load greater than or equal to In accordance with its associated single largest post-accident load, and: the Surveillance

a. Following load rejection, the speed is < 75% of Frequency Control the difference between nominal speed and the Program overspeed trip setpoint; and

b. Following load rejection, the voltage is

> 3740 V and < 4580 V.

Farley Units 1 and 2 3.8.1-8 Amendment No.r5jI (Unit 1)

Amendment No. F1 (Unit 2)

FnrJngi re '2 tn N1 1 3R.S However, portions of the surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety AC Sources - Operating of the plant is maintained or enhanced. 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.9 --------------------------- N TES---------------

1. All DG starts may e preceded by an engine prelube period.

2. This Surveillance s l11not be performed in MODE 1, 2, 3, or 4normally Verify on an actual or simulated loss of offsite power In accordance with signal: the Surveillance Frequency Control

a. De-energization of emergency buses; Program

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in < 12 seconds,

2. energizes auto-connected shutdown loads through automatic load sequencer,

3. maintains steady state voltage

> 3740 V and* 4580 V,

4. maintains steady state frequency

> 58.8 Hz and *61.2 Hz, and

5. supplies permanently connected and auto-connected shutdown loads for

> 5 minutes.

Farley Units 1 and 2 3.8.1-9 Amendment No. 6 (Unit 1)

Amendment No. f (Unit 2)

NL-14-1385, Basis for Proposed Changes. Part 1 of 2

Text

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