Text

Response to Request for Additional Information Regarding Multiple Technical Specifications Changes
	ML15271A223
Person / Time
Site:	Farley
Issue date:	09/28/2015
From:	Pierce C; Southern Co, Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-15-1814
	Download: ML15271A223 (46)
	v • d • e

{{#Wiki_filter:Charles A. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc. 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 SOUTHERN << \ Fax 205.992.7601 NUCLEA A SOUTHERN COMPANY SEP 2 8 2015 Docket Nos.: 50-348 NL-15-1814 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding Multiple Technical Specifications Changes Ladies and Gentlemen: By letter dated November 24, 2014, (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14335A623, ML14335A624, ML14335A629) Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) to adopt various previously approved Technical Specifications Task Force (TSTF) Travelers for Joseph M. Farley Nuclear Plant (FNP). By letter .dated August 14, 2015, the Nuclear Regulatory Commission (NRC) sent SNC a request for additional information (RAI). Enclosure 1 provides the SNC response to the NRC RAI. SNC proposes to revise the Technical Specifications and Bases mark-up and clean-typed pages submitted as part of the LAR, as provided in Enclosures 2 and 3 of this letter. This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

U.S. Nuclear Regulatory Commission NL-15-1814 Page2 Mr. C. R. Pierce states he is the Regulatory Affairs Director for Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the -Rest~"J:p;* facts set forth in this letter are true. C. R. Pierce Regulatory Affairs Director CRP/JMC/ dayof1/)~ ,2015. My commission expires: (o-f- 2 0 11

Enclosures:

1. SNC Response to NRC RAI

2. Revised Technical Specifications and Bases Markup Pages

3. Revised Technical Specifications Clean-Typed Pages cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President- Farley Mr. M. D. Mejer, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Ms. B. L. Taylor, Regulatory Affairs Manager- Farley RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. L. D. Wert, Regional Administrator (Acting)

Mr. S. A. Williams, NRR Project Manager- Farley Mr. P. K. Niebaum, Senior Resident Inspector- Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer

Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding Multiple Technical Specifications Changes Enclosure 1 SNC Response to NRC RAI

Enclosure 1 to NL-15-1814 SNC Response to NRC RAJ By letter dated November 24, 2014, (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14335A623, ML14335A624,

ML14335A629) Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) to adopt various previously approved Technical Specifications Task Force (TSTF) Travelers for Joseph M. Farley Nuclear Plant (FNP). By letter dated August 14, 2015, the Nuclear Regulatory Commission (NRC) sent SNC a request for additional information (RAJ). This enclosure provides the SNC response to the NRC RAI.

Request for Additional Information #1, TSTF-247: This RAJ is with respect to TSTF-247 and TS 3.4.11, Condition F. The proposed changes provide separate Condition entry for each power operated relfef valve (PORV) and each block valve. In explaining the difference between the plant-specific justification and the approved traveler justification, the licensee states, in part, Condition F is modified to apply when both 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 hours, 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." Clarify the justification regarding the fact that the PORVs are not currently credited as part of the FNP LTOP strategy and that the applicability of FNP limiting condition for operation (LCO) 3.4.11 is "MODES 1, 2 and 3," and not "LTOP" conditions; LTOP does not appear relevant to this LCO. SNC Response to RAJ #1, TSTF-247: The changes in TSTF-247-A provide separate condition entry for each Pressurizer Power Operated Relief Valve (PORV) and block valve. As described in TSTF-247-A, the existing LCO 3.4.11 Conditions allow separate condition entry for each PORV, and the Conditions and Required Actions provide appropriate compensatory measures for separate condition entry. 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. The proposed modification of the Actions Note for TS 3.4.11 to allow separate condition entry for each block valve is not based on low temperature overpressure protection (LTOP) considerations, and is applicable for FNP. FNP TS 3.4.11, Required Action F.1 requires that the PORVs be placed in manual control within 1 hour if both block valves are inoperable. TSTF-247-A provides a basis for eliminating TS 3.4.11, Required Action F.1 that is based on the need to preserve the PORV safety function for LTOP. Based on the fact that the basis for elimination of Required Action F.1 is based on LTOP considerations E1 - 1

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI which are not applicable to FNP, the changes to Required Action F.1 are not adopted. The changes in TSTF-247-A eliminate Required Action F.3 (Restore remaining block valve(s} to OPERABLE status} because, with separate condition entry for each block valve, it is not needed. The proposed changes toTS 3.4.11 resulting from adoption of TSTF-247-A are revised to reinstate Required Action F.1 and conforming changes are made to the TS 3.4.11 Bases for Condition F. Revised TS and Bases markup pages and clean-typed TS pages reflecting these changes are provided in Enclosures 2 and 3. Request for Additional Information #2, TSTF-283: This AAI is with respect to TSTF-283, and SA 3.8.4. 7 and SA 3.8.4.8. In adopting TSTF-283 on modifying Mode restriction notes, the applicant does not request to adopt the modification the surveillances related to the battery: SA 3.8.4.7 on Battery capacity, and SA 3.8.4.8 on Battery discharge test. The reason for not requesting these changes is neither discussed nor justified. Justify not adopting the appropriate TSTF-283 changes on modifying Mode restriction notes to SA 3.8.4.7 and SA 3.8.4.8. SNC Response to RAI #2, TSTF-283: TS and Bases markups have been revised to include the TSTF-283 changes for SA 3.8.4.7 and SA 3.8.4.8, with the following difference from the approved Traveler. ~ TSTF-283-A, Rev. 3 provides supplemental BASES text for SA 3.8.4.7 and SA 3.8.4.8 which states,

        This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY ... provided an assessment determines plant safety is maintained or enhanced."

and,

        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 in MODE 1 or 2."

The MODE restrictions in Note 2 of Farley SA 3.8.4.7 and SA 3.8.4.8 on normally performing the Surveillance are applicable in MODES 1, 2, 3, or 4. However, the applicability of the allowance provided in the supplemental BASES text from TSTF-283-A is limited to "MODES 1 or 2." E1- 2

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI The intent of TSTF-283-A is to allow performance of surveillance testing for the purposes of reestablishing OPERABILITY provided an assessment of the effects of performance in these MODES compared to the effects of a plant shutdown and startup demonstrate a safety benefit or safety neutral situation. The supplemental BASES text provided in TSTF-283-A is therefore expanded from "MODES 1 or 2" to "MODES 1, 2, 3, or 4," which is consistent with the MODE restrictions in Note 2 of Farley SR 3.8.4.7 and SR 3.8.4.8, and the intent of TSTF-283-A, Rev. 3. The proposed changes to the Bases for SR 3.8.4. 7 and SR 3.8.4.8 will allow performance of the testing specified by these SRs in all Modes of operation. This will help to reduce the complexity of coordinating work and testing activities during refueling outages and could potentially reduce outage critical path time. The change will also maximize flexibility in responding to an event during shutdown when other engineered safety features (ESF) equipment may be out of service. In addition, this change could potentially avoid a plant shutdown if maintenance (planned or unplanned) performed in MODES 1, 2, 3, or 4 results in the need to perform the surveillance to demonstrate operability. Revised TS and Bases markup pages and clean-typed TS pages reflecting these changes are provided in Enclosures 2 and 3. Request for Additional Information #3, TSTF-284: This RAI is with respect to TSTF-284 and the Bases to SR 3.4.11.1. In adopting TSTF-284 on "Met vs Performed" clarifications, the first part of Westinghouse Owner's Group (WOG) Insertion A is not included in the proposed change to Bases for SR 3.4.11.1. Missing is the sentence that states, "Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable." Include the complete Insertion A to the SR 3.4.11.1 Bases. SNC Response to RAI #3, TSTF-284: The first part of WOG Insertion A has been added to the Bases for SR 3.4.11.1. A revised Bases markup page reflecting this change is provided in Enclosure 2. Request for Additional Information #4, TSTF-308: This RAt is with respect to TSTF-308 and TS 5.5.4.e. The two sentences that replace the text in TS 5.5.4.e are accurate and align with NUREG 1431. However, the second sentence should end with a semi-colon instead of a period to ensure the correct formatting and to align the text with the STS. E1- 3

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI SNC Response to RAI #4, TSTF-308: The proposed changes for TS 5.5.4.e are revised to end the second sentence with a semi-colon, and not a period." A revised TS markup page and clean-typed TS page reflecting this change are provided in Enclosures 2 and 3. Request for Additional Information #5, TSTF-312: This RAI is with respect to TSTF-312 and LCO 3.9.3. In explaining the difference between TSTF-312-A requirements and the proposed changes to FNP TS 3.9.3 and its associated bases, the licensee states that LCO 3.9.3.b was previously amended to allow the personnel and emergency personnel airlocks to remain open during CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment, and the scope of this previous amendment (FNP Amendment 165l157, dated September 30, 2004) overlaps the scope of TSTF 312-A, resulting in the statement of LCO 3.9.4 and its associated bases being different from those presented in TSTF-312-A. The NRC staff reviewed the referenced Amendment 165/157, and noted that this Amendment addresses the allowance for the open equipment hatch, not the open air locks. The NRC staff requests that the licensee identify the FNP license amendment that approved allowing the personnel airlocks to remain open during CORE ALTEAATIONS and during movement of irradiated fuel assemblies within the containment. SNC Response to RAI #5, TSTF-312: Changes to FNP TS 3.9.3, "Containment Penetrations," to allow the equipment hatch to be open during core alterations and/or during movement of irradiated fuel assemblies within containment were approved by the NRC as Amendment 165/157 in a letter from Sean E. Peters (NRC) to L. M. Stinson (SNC), dated September 30, 2004 (ADAMS Accession No. ML042860246). Changes to FNP TS 3.9.3 to allow the containment personnel air locks that provide direct access from the containment atmosphere to the auxiliary building to be open during refueling activities were approved by the NRC as Amendment 178/171 in a letter from Karl D Feintuch (NRC) to J. A. Johnson (SNC), dated September 29, 2006 (ADAMS Accession No. ML082730007). The changes in amendment 178/171 were based on Technical Specifications Task Force (TSTF) Traveler TSTF-68-A, Revision 2. As documented in the Safety Evaluation Report (SEA) for Amendment 165/157, the staff's evaluation of the change to allow the equipment hatch to be open during core alterations and/or during movement of irradiated fuel assemblies within containment reviewed and found acceptable the assumptions, inputs, and methods used in the fuel handling accident (FHA) consequence analysis. E1- 4

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI A revised FHA dose consequence analysis, with the containment equipment hatch and personnel air locks open during a postulated FHA inside containment, was provided in support of the TS 3.9.3 changes that were made in Amendment 178/171. As described in the SEA for Amendment 178/171 , the NRC staff performed a confirmatory FHA dose consequence calculation for these changes and found the results to be within the dose criteria specified in 10 CFR 100.11 and GDC 19 of 10 CFR Part 50, Appendix A. Request for AdditionaJJnformation #6, TSTF-315: This RAI is with respect to TSTF-315 and 3.1.8. The text added to the LCO statement in TS 3.1.8 and the corresponding Bases reads" ... may be reduced to 3, provided: ... " The text in the STS reads " ... may be reduced to 3 required channels, provided: ... " This is to ensure the accuracy and completeness of the TS and the Bases and to align the TS and Bases with the STS. The NRC staff requests that the licensee justify the deviation from the STS, NUREG 1431. SNC Response to RAJ #6, TSTF-315: The proposed changes for TS LCO 3.1.8 and the LCO 3.1.8 Bases are revised to add "required channels." Revised TS and Bases markup pages and clean-typed TS pages reflecting these changes are provided in Enclosures 2 and 3. Request for AdditionaJJnformation #7, TSTF-343: This RAI is with respect to TSTF-343 and TS 5.5.17. In explaining the difference between TSTF-343-A changes in STS requirements and the proposed changes to FNP TS 5.5.17 requirements, the licensee states that the changes identified for FNP TS 5.5.6, "Pre-stressed Concrete Containment Tendon Surveillance Program," and conforming changes to the TS Bases for SR 3.6.1.2 and a reference to RG 1.35 are not adopted because those changes are already reflected in the current FNP TS and Bases. The applicable FNP license amendment numbers for those changes were not stated in the application. The NRC staff requests that the licensee identify the applicable FNP license amendments that approved those changes for incorporation into the FNP TS. SNC Response to RAJ #7, TSTF-343: Changes corresponding to those identified in TSTF-343-A, Rev. 1 forTS 5.5.6, "Pre-stressed Concrete Containment Tendon Surveillance Program," the TS Bases for SR 3.6.1.2, and the Bases References forTS 3.6.1 were approved by the NRC under license amendment 172/165 in a letter from Robert. E. -Martin (NRC) to L. M. Stinson (SNC), dated April14, 2006 (ADAMS Accession Nos. ML060830368 and ML061150446). E1- 5

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI Request for Additional Information #8, TSTF-349: This RAI is with respect to TSTF-349 and LCO 3.9.6 Bases statement. LCO 3.9.6 Bases statement is revised to add a second Note permitting all RHR pumps to be de-energized for no more than 15 minutes when switching from one RHR train to another. The NRC staff noted the following in the revised Bases discussion of the new LCO Note: In the second sentence of the added paragraph, parentheses are placed around "and the core outlet temperature is limited to > 10 degrees F below saturation temperature" for no apparent reason. In accordance with the Writer's Guide, parentheses are used to indicate clarifying details for the preceding text. The NRC staff requests that the licensee explain the use of parentheses in this case. SNC Response to RAI #8, TSTF-349: The LCO 3.9.5 Bases are revised to remove the braces surrounding the expression " ... and the core outlet temperature is limited to> 10 degrees F below saturation temperature." A revised Bases markup page is provided in Enclosure 2. Request for Additional Information #9, TSTF-371: This RAI is with respect to TSTF-371, and SR 3.3.1.2 and SR 3.3.1.3. In TSTF-371 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. The technical part of TSTF-371 was approved on a plant-specific basis for FNP, Units 1 and 2, TS SR 3.3.1.2 and SR 3.3.1.3 in 1999. The present LAR proposes to revise the presentation of these Surveillance Requirements to match the final version of TSTF-371-A approved by NRC in 2002.

This LAR involves an editorial change in presentation of the requirement to adjust [increase] Nuclear Instrument System (NIS) power range

[neutron flux] channel [output] to match calorimetric heat balance calculation [results] if the calorimetric heat balance calculation [results] exceed the power range [neutron flux] channel output by more than +2% RTP. The requirement is moved from surveillance column Note 1 to the surveillance statement, consistent with the TSTF change. The change deletes SR 3.3.1.2 Note 1, which states: E1- 6

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI "Adjust NIS channel if calorimetric calculated power exceeds NIS indicated power by more than +2% RTP." The change modifies the surveillance statement of SA 3.3.1.2 as follows:

       "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%."

RAI question 1: Request licensee to insert "RTP" at end of surveillance statement.

This LAR involves an editorial change in presentation of the requirement to adjust [increase or decrease] Nuclear Instrument System (NIS) power range [neutron flux] channel output if the absolute difference between the NIS AXIAL FLUX DIFFERENCE (AFD) and the incore [neutron flux] detector measurements [of AFD] are greater than or equal to 3% RTP. The requirement is moved from surveillance column Note 1 to the surveillance statement, consistent with the TSTF change.

The change deletes SA 3.3.1.3 Note 1, which states:

           "Adjust NIS channel if absolute difference is 2: 3%."

The change modifies the surveillance statement of SA 3.3.1.3 as follows:

           "Compare results of the incore detector measurements to Nuclear Instrumentation System (NIS) AFD. Adjust NIS channel if difference is 2: 3%."

RAI question 2: Request licensee to insert "RTP" at end of surveillance statement; and insert "absolute" before "difference" in second sentence. SNC Response to RAI #9, TSTF-371: SA 3.3.1.2 is revised to insert "RTP" at end of surveillance statement. The TS and Bases for SA 3.3.1.3 is revised to insert "RTP" at end of surveillance statement, and insert "absolute" before "difference" in the second sentence. The Bases for SA 3.3.1.3 are also revised to add "RTP" in the surveillance description. Revised TS and Bases markup page and clean-typed TS pages reflecting these changes are provided in Enclosures 2 and 3. E1 -7

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI Request for Additional Information #1 0, Changes related to ISTS Adoption

#1:

When the licensee for FNP, Units 1 and 2, converted the plant-specific custom TS to plant-specific improved TS in about 1997, it elected not to increase the 24-hour Completion Time to restore to operable status an inoperable channel of the P-4 (Reactor Trip) ESFAS interlock, improved TS 3.3.2 Function 7.b, to a Completion Time of 48 hours, which had been the requirement in Westinghouse STS since 1981. In this LAR, the licensee proposes to adopt the 48 hour Completion Time by changing the specified Condition for this Function in Table 3.3.2-1 from Condition C to Condition F. The justification for the change presented in LAR letter Enclosure 1 is essentially to achieve consistency with the improved STS. The licensee is requested to provide a technical safety basis for this change that is consistent with the FNP, Unit 1 and 2 licensing basis, including the design of the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) instrumentation, the FSAR Chapter 15 safety analyses, and the effects of an inoperable P-4 channel on the affected ESFAS Functions and ESF systems in the event of a reactor trip. SNC Response to RAI #1 0. Changes related to ISTS Adoption #1: The proposed change to FNP TS Table 3.3.2-1, Function 7.b, "Reactor Trip, P-4," increases the Completion Time to restore an inoperable channel of the P-4 (Reactor Trip) Engineered Safety Feature Actuation System (ESFAS) interlock to operable status from 24 hours (Condition C) to 48 hours (Condition F). The corresponding ESFAS interlock requirements identified in FNP TS Table 3.3.2-1, Function 7.b, are presented in Improved Standard Technical Specification (ISTS) Table 3.3.2-1, Function 8.a. The Reactor Trip System (RTS) initiates a trip of the main turbine when a reactor trip (P-4) is generated. The P-4 interlock is enabled when a reactor trip breaker and its associated bypass breaker are open. The P-4 interlock does not provide a safety design function for the RTS, and does not have the ability to affect RTS operation or reliability. The P-4, Reactor Trip interlock provides the following ESFAS control and safety design functions: Control

Block steam dump control via load rejection controller;
Arm steam dump control for tripping and/or modulation of dump valves via turbine trip controller; and
Isolate main feedwater (MFW) with coincident low Tavg.

Safezy

Prevent auto reactuation of Sl after a manual reset of Sl;
Trip the main turbine; E1- 8

Enclosure 1 to NL-15-1814 SNC Response to.NRC RAI

Reset high steam flow setpoint to no-load value; and
Prevent opening of the MFW isolation valves if they were closed on Safety Injection (SI) or Steam Generator Water Level- High High.

As described in the Applicable Safety Analysis for FNP Bases 3.3.2, the Reactor Trip, P-4 interlock is assumed in the large break LOCA safety analyses since the block of automatic safety injection (SI) signals is required to support long-term Emergency Core Cooling System (ECCS) operation and realignment to the post-LOCA recirculation mode after the initial phase of Sl injection is complete. Manual realignment of the ECCS system to support long-term ECCS operation following a large break LOCA is a typical strategy for Westinghouse pressurized water reactors, and is not unique safety function requirement for FNP. The safety design functions of the FNP P-4 interlock are consistent with the safety function description provided for the Reactor Trip, P-4 interlock in the ISTS Applicable Safety Analysis for Bases 3.3.2, which states that none of the control or safety functions associated with the P-4 interlock serve a mitigation function in the licensing basis safety analyses, or are required to show that the licensing basis safety analysis acceptance criteria are not exceeded. The proposed 48 hour Completion Time is consistent with the Completion Time identified for the Reactor Trip, P-4 interlock that is provided in Appendix A, "Proposed Changes to the Standard Technical Specifications," of WCAP-10271-P-A, Supplement 2, Rev. 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," June 1990. Adoption of a 48 hour Completion Time (Condition F) for the Reactor Trip, P-4 interlock in FNP TS Table 3.3.2-1, Function 7.b, is also consistent with the Completion Time that has been reflected in ISTS (NUREG 1431) for this function since Rev. 0. WCAP-10271-P-A, Supplement 2, Rev. 1 is identified as Reference 10 in the FNP Bases forTS 3.3.2. Request for Additional Information #11, Voatle Consistency Change #1: The "Vogtle Consistency Change #1" to revise the Completion Time from 4 hours to 8 hours for Required Action A.1, to restore seal injection flow, is not consistent with the STS, nor is the change adequately justified. Similarly the time provide by the NOTE to SR 3.5.5.1 allowing 8 hours rather than 4 hours for flow to stabilize, is not consistent with the STS nor is it adequately justified. Adequately justify this proposed change, and propose this generic Technical Specification change through the Technical Specification Task Force for review and approval. SNC Response to RAI #11, Voqtle Consistency Change #1: The requested Completion Time change identified in Vogtle Consistency Change

1 for FNP SR 3.5.5.1 is withdrawn.

E1- 9

Enclosure 1 to NL-15-1814 SNC Response to NRC RAI Discrepancy noted during the review of TSTF 315 but not related to TSTF315: While reviewing the adoption of TSTF 315, the NRC staff noted a discrepancy in the Bases of TS 3.1 .8. The LCO statement for TS 3.1.8 lists three conditions that must be met in order to perform Physics Tests. Statement "a" reads THERMAL POWER is s5% RTP;". Statement "c" reads "RCS lowest loop average temperature is <::531 °F." While reviewing the mark-up pages provided for the Bases of TS 3.1.8, the NRC staff noted that those same statements do not read the same as in the TS. Statement "a" reads THERMAL POWER is =5% RTP; and". Statement "c" reads "RCS lowest loop average temperature is =531 °F." The Bases replace the inequality signs with equal signs. This is inconsistent with the LCO statements forTS 3.1.8, and is also inconsistent with the STS, NUREG 1431. Justify the inconsistency. SNC Response to Discrepancy noted during the review of TSTF 315 but not related to TSTF315: The Bases markup for LCO 3.1.8 is been revised to reflect the following changes for statements "a" and "c":

THERMAL POWER is S5% RTP; and
RCS lowest loop average temperature is <::531 °F.

A revised Bases markup page reflecting these changes is provided in . E1 - 10

Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding Multiple Technical Specifications Changes Enclosure 2 Revised Technical Specifications and Bases Markup Pages to NL-15-1814 Revised Technical Specifications and Bases Markup Pages Pressurizer PORVs 3.4.11 jRAI #1 ACTIONS CONDITIOry-!Two I REQUIRED ACTION COMPLETION TIME 1/ F. I Mere thaR eRelfjlock~ F.1 Place associated 1 hour inoperable. PORVs in manual control.

                      !valves AND F.2            Restore one block valve                       2 hours to OPERABLE status.
                                                  -ANa-r . .;1 I"'\ I::::>> LUI._. 1 ._.111a11 Ill 1~

I'-

                                                                                                                          ..IUUI<>

sleek \*alote te 6PERABl:E stet1:1s. G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time of Condition F not AND met. G.2 Be in MODE4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 !Performed D NOTES--------

1. Not required to b~~ith block valve closed in accordance with the Requir~lt:!llll\:cJetw:iamn~o~fLI,--____,,....,.....,_L~-.

I CeAaitieR B er EJ Actions of this LCO.I

2. , Not required to be performed prior to entry into M6DE3.

I ITSTF-284

                              --------------- ----------fr-                                                                                     I Perform a complete cycle of each block valve.                                              In accordance with
                                !Only required to be performed in MODES                                                 the Surveillance Frequency Control 11 and 2.                                                                                Program Farley Units 1 and 2                                     3.4.11-3                                     Amendment No. ~ (Unit 1)

Amendment No. ~ (Unit 2) E2-1 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages Pressurizer PORVs B 3.4.11 IRAI #1 BASES ACTIONS D.1 and D.2 (continued) If the Required Action of Condition A, B, or C is not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE. E.1. E.2. E.3. and E.4 If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the Completion Time of 1 hour or isolate the flow path by closing and removing the power to the associated block valves. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time and provides the operator time to correct the situation. If one PORV is restored and one PORV remains inoperable, then the plant will be in Condition B with the time clock started at the original declaration of having two PORVs inoperable. If no PORVs are restored within the Completion Time, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE.

                      ~                                                  two block valves are If                                moperable, it is necessary to either restore the block valves within the Completion Time of 1 hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hoursIaREI FestoFe tl'te FeA'IBiRiAg bloeiE vahre I l'tt*ithin 72 hour§. The Completion Times are reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation.

(continued) Farley Units 1 and 2 B 3.4.11-6 Revision~ E2-2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages DC Sources- Operating 3.8.4 jRAI#2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.6 ----------*--*---NOT E-----*------------ This Surveillance may be performed in MODE 1, 2, 3, 4, 5, or 6 provided spare or redundant charger(s) placed in service are within surveillance frequency to maintain DC subsystem(s) OPERABLE. Verify each required Auxiliary Building battery In accordance with charger supplies ~ 536 amps at ~ 125 V for ~ 4 hours the Surveillance and each required SWIS battery charger supplies Frequency Control

                              ~ 3 amps at~ 125 V for~ 4 hours.                            Program SR 3.8.4.7               -----------NOTES----------

1. The performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4. 7 once per 60 months.

2. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test at any time. ~

3. This Surveillance shall not 'tfe performed for the Auxiliary Building batteries in MODE 1, 2, 3, or4.

Verify batte capacity is adequate to supply, and In accordance with maintain in ERABLE status, the required the Surveillance emergency lo ds for the design load profile described Frequency Control in the Final Sa ety Analysis Report, Section 8.3.2, by Program subjecting the attery to a service test. However, portions of the Surveillance may be

                                                 \ _ performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced .

Farley Units 1 and 2 3.8.4-3 Amendment No. 185 (Unit 1) Amendment No. 180 (Unit 2) E2-3

   ~ However, portions of the Surveillance may be performed to reestablish OPERABILITY provided                        DC Sources- Operating an assessment determines the safety of the plant is                                      3.8.4 maintained or enhanced .

IRAI#2 SURVEILLANCE REQUIREMENTS SURVEILLANCE ;-{normally _I FREQUENCY SR 3.8.4.8 This Surveillance shall ~o~~- perfonned for the _'I' Auxiliary Building batteries in MODE 1, 2, 3, or 4. Verify battery capacity is ;;:: 80% of the manufacturer's In accordance with rating when subjected to a performance discharge the Surveillance test or a modified performance discharge test. Frequency Control Program AND 18 months when battery shows degradation or has reached 85% of expected life or 17 years, whichever comes first Farley Units 1 and 2 3.8.4-4 Amendment No. 185 (Unit 1) Amendment No. 180 (Unit 2) E2-4

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages DC Sources- Operating B 3.8.4 jRAI#2 BASES SURVEILLANCE SR 3.8.4.6 (continued) REQUIREMENTS (continued) The spare or redundant battery and/or charger must be within the 18 month surveillance frequency to maintain the DC subsystem(s) to which they are aligned OPERABLE. This operational flexibility maintains TS OPERABILITY of the applicable battery and DC train while testing the normally aligned charger. SR 3.8.4.7 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (design load profile) of the DC electrical power system. The discharge rate and test length should correspond to the design load profile requirements as specified in Reference 4. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is modified by three Notes. Note 1 allows the performance of a performance discharge test in lieu of a service test once per 60 months. Note 2 allows the performance of a modified performance discharge test in lieu of a service test at any time. The modified performance discharge test is a simulated duty cycle consisting of just two rates: the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelop the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery

                                . terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test. (continued) Farley Units 1 and 2 B 3.8.4-9 Revision 59 E2-5

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages DC Sources- Operating B 3.8.4 jRAI#2 BASES SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS The reason for Note 3 is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems. INSERT- Bases SR 3.8.4 SR 3.8.4.8 A battery performance discharge test is a test of constant current capacity of a battery, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage. A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8. The modified performance discharge test may be used to satisfy SR 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 at any time. The performance discharge test may be used to satisfy 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 once per 60 months. The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 9). This reference recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life or 17 years, whichever comes first, the Surveillance Frequency is reduced to 18 months. Degradation is indicated, according to IEEE-450 (Ref. 9), when the battery capacity drops by mar~ than 10% relative to its capacity on the previous performance test or when it is > 10% below the manufacturer's rating. This SR is modified by a Note. The reason for the Note is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems. INSERT- Bases SR 3.8.4 (continued) Farley Units 1 and 2 B 3.8.4-10 Revision 59 E2-6 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages IRAI#2 INSERT- Bases SR 3.8.4 This restriction from normally performing the Surveillance in MODES 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance 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 partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial 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 portions of the Surveillance are performed in MODES 1, 2, 3, or 4. Risk insights or deterministic methods may be used for this assessment. E2-7

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages Pressurizer PORVs B 3.4.11 lRAI#3 BASES ACTIONS G.1 and G.2 (continued) If the Required Actions of Condition F are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE. SURVEILLAN CE SR 3.4.11.1 REQUIREMEN TS Opening the block valve in 1-- - -- -- - -- -- - -- - ------- this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable. In accordance with Reference 3, administrative controls require this test to be performed in MODE 3 or 4 to adequately simulate opening temperature and pressure effects on PORV operation . (continued) Farley Units 1 and 2 B 3.4.11-7 Revision ~ E2-8

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages Programs and Manuals 5.5 jRAI#4 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration stated in 10 CFR 20, Appendix B (to paragraphs 20.1001-20.2401),

Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination ef eumulath," ana prejeetea aese ooRtFieutieAs from radieaetive effluents for the etn=rent calendar quarter and current calendar year in aooerdanee witl:l tl:le metf:leaelegy eRa parameters iR the ODCM at least e*t'ef)* 31 aays;

f. Limitations on the functif nal capability and use of the liquid and gaseous effluent treatment syste(ns to ensure that appropriate portions of these systems are used to r~ uce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose c lmmitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the do e rate resulting from radioactive material released in gaseous effluents to reas at and beyond the site boundary as follows:

1. For noble gase : Less than or equal to a dose rate of 500 mrem/year to the total boqy and less than or equal to a dose rate of 3000 mrem/year to Jhe skin, and

2. For lodine-1 31 , lodine-133, tritium, and for all radionuclides in particulate f~im with half lives greater than 8 days: Less than or equal to a dose ratf of 1500 mrem/year to any organ.

h. Limitations on thj: annual and quarterly air doses resulting from noble gases released n gaseous effluents from each unit to areas beyond the site boundary, 1onforming to 10 CFR 50, Appendix I; Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at (continued) least every 31 days; Farley Units 1 and 2 5.5-3 Amendment No. Hftl (Unit 1)

Amendment No. ~ (Unit 2) E2-9

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Mai'Xup Pages PHYSICS TESTS Exceptions-MODE 2 3.1.8 IRAI#6 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)"; LCO 3.1.4, "Rod Group Alignment Limits"; and the number of LCO 3.1.5, "Shutdown Bank Insertion Limits"; LCO 3.1.6, "Control Bank Insertion Limits"; and required channels for LCO 3.4.2, "RCS Minimum Temperature for Criticality" LCO 3.3.1, "RTS Instrumentation ,"

                                                        ~----------------------__,

may be suspended, provided: Functions 2, 3, and 17.e, may be reduced to 3 required channels

a. THERMAL POWER is s 5% RTP;

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

c. RCS lowest loop average temperature is C!: 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 hour 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. ~(Unit 1) Amendment No. ~(Unit 2) E2 -10

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages PHYSICS TESTS Exceptions- MODE 2 B 3.1.8 IRAI#6 BASES APPLICABLE not violated. When one or more of the requirements specified in SAFETY ANALYSES LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, (continued) LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteri.a are preserved as long as the power level is limited to = 5% RTP, the reactor coolant temperature is kept

                                           = 531°F, and SDM is within the limits provided in the COLR.

The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. PHYSICS TESTS meet the criteria for inclusion in the Technical Specifications, since the components and process variable LCOs suspended during PHYSICS TESTS meet Criteria 1, 2, and 3 of 10 CFR 50.36 (c)(2)(ii) . Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity. LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits Operation beyond specified limits is permitted for the purpose of erforming PHYSICS TESTS and poses no threat to fuel integrity provided the SRs are met. One Power Range Neutron Flux The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, Channel may be and LCO 3.4.2 may be suspended ring the performance of bypassed , reducing PHYSICS TESTS provided: ,....a_n_d-th_e_n_u_m_b,_e-r-o""""f- -, the number of required channels for required channels a. THERMAL POWER is = 5% RTP; and LCO 3.3.1, "RTS . from 4 to 3. Instrumentation,"

b. SDM is within the limits provided in the COLR; and Functions 2, 3, 6, and 17.e, may be reduced

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

to 3 required channels Farley Units 1 and 2 B 3.1.8-5 Revision~ E2 -11

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages RHR and Coolant Circulation- Low Water Level B 3.9.5 IRAI#8 BASES LCO An OPERABLE RHR loop consists of an RHR pump, a heat (continued) exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. ritwo Notes. The first Note I The LCO requirements are modified by la Nate whienlprovides an exception to the requirements for one RHR loop to be OPERABLE and one RHR loop to be in operation. This exception is necessary to ensure the RHR System may be realigned as necessary for up to 2 hours to perform the required surveillance testing necessary to verify the RHR System performance in the ECCS injection mode of operation . APPLICABILITY Wwo RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the op of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level ~ 23 ft are located in LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level." ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action The second Note permits the shall be immediately initiated and continued until the RHR loop is RHR pumps to be de- restored to OPERABLE status and to operation or until ~ 23 ft of energized for ~ 15 minutes water level is established above the reactor vessel flange. When the when switching from one water level is ~ 23 ft above the reactor vessel flange, the Applicability train to another. The changes to that of LCO 3.9.4, and only one RHR loop is required to circumstances for stopping be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions. both RHR pumps are to be limited to situations when the outage time is short and the t-' core outlet temperature is If no RHR loop is in operation, there will be no forced circulation to limited to > 10 degrees F provide mixing to establish uniform boron concentrations. Reduced below saturation boron concentrations can occur by the addition of water with a lower temperature. The Note boron concentration than the required boron concentration specified prohibits boron. dilution or in the COLR.,. Therefore, actions that could result in the addition of draining operations when water to the RCS with a boron concentration less than the required RHR forced flow is stopped . boron concentration specified in the COLR must be suspended immediately. (continued) Farley Units 1 and 2 B 3.9.5-2 Revision~ E2 -12

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS IRAI #9

            ------------------ -------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 hour after THERMAL POWER is< P-6. Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 NOTES

                                              ==~~==~~:1= 1 1.

power range channel output Adjust power I range channel output if ~ Not required to be performed until 24 hours calorimetric heat balance after THERMAL POWER is ~ 15% RTP. calculation results exceed power range Compare re\~Jits o-fcal~rlme~r~~h~~balance______ In accordance with channel output by more than +2% RTP. calculation to ,

                                                 ,.., .... ;-;.,l  -   lll:::>ll UIIII:OIIlCUIUII ;;:)y:::>Lt:lll \ l'tl;;:) Jl  the Surveillance Frequency Control Program SR 3.3.1.3                                              NOTES I ~.     ~~::.t NIS el'laAFtel if aeselute aiffereFtee is                            I
                   ~[!]                       Not required to be performed until 7 days after THERMAL POWER is~ 50% RTP.
                  ~~                          Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the in core detector In accordance with measurements t AFD. ~ Adjust NIS channel if the Surveillance absolute difference is Frequency Control

     !Nuclear Instrumentation System (NIS)                                                                                      Program 2:3% RTP.

Farley Units 1 and 2 3.3.1-9 Amendment No. h esl (Unit 1) Amendment No. ~ (Unit 2) E2-13

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages RTS Instrumentation B 3.3.1 jRAI #9 BASES SURVEILLANCE SR 3.3.1.3 REQUIREMENTS (continued) SR 3.3.1.3 compares the incor ystem to the NIS channel output. If the absolute difference is ~ 3% the NIS channel is still OPERABLE, but it must be adjusted. j V V llwl Lllw \IIICIIIIIwl I~ UUL~IUw Lllw oJ IU CUIUVVCIII\Iw The excore NIS channel ass~FRed in the setpeint ~noertaiAty eale1:1latiefl, tJ:te ei=IBFIFiel m1:1st be shall be adjusted if the

                                            .       t:
                                                       ..        .    ...1\            *             "'          _..

absolute difference between the incore and _7! If the NIS channel cannot be properly adjusted, the channel is declared excore AFD is;;::: 3% RTP. inoperable. This Surveillance is performed to periodically verify the and excore AFD is ~ S% . Note clarifies that the Surveillance is required only if reactor power is ~ 50% RTP and that 7 days are allowed for performing the Surveillance and channel adjustment, if necessary, after reaching 50% RTP. A power level of ~ 50% RTP is consistent with the requirements of SR 3.3.1.9. Note llows SR 3.3.1.9 to be performed in lieu of SR 3.3.1.3, since S 3.3.1.9 calibrates (i.e., requires channel adjustment) the excore c annals to the incore channels, it envelopes the performance of 3.3.1.3. 2 For each operating cycle, the initial channel normalization is performed under SR 3.3.1.9. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT. This test shall verify OPERABILITY by actuation of the end devices. The RTB test shall include separate verification of the undervoltage trip via the Reactor Protection System and the local manual shunt trip mechanism. The bypass breaker test shall include a local manual shunt trip and local manual undervoltage trip. A Note has been added to indicate that this test must be performed on a bypass breaker prior to placing it in service. The independent test of undervoltage and shunt (continued) Farley Units 1 and 2 B 3.3.1-52 Revision lsz l E2-14

Enclosure 2 to NL-15-1814 Revised Technical Specifications and Bases Markup Pages PHYSICS TESTS Exceptions- MODE 2 B 3.1.8 TSTF-315 Discrepancy Change BASES APPLICABLE not violated. When one or more of the requirements specified in SAFETY ANALYSES LCO 3.1.3, "Moderator Temperature Coefficient {MTC)," LCO 3.1.4, {continued) LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power

level is limited to 5% RTP, the reactor coolant temperature is kept

                                            = 531°F, and SDM is within the limits provided in the COLR..

The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components {control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. PHYSICS TESTS meet the .criteria for inclusion in the Technical Specifications, since the components and process variable LCOs suspended during PHYSICS TESTS meet Criteria 1, 2, and 3 of 10 CFR 50.36 {c){2){ii). Reference 6 allows special test exceptions {STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity. LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits Operation beyond specified limits is permitted for the purpose of erforming PHYSICS TESTS and poses no threat to fuel integrity provided the SRs are met. One Power Range Neutron Flux The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, Channel may be and LCO 3.4.2 may be suspended ring the performance of bypassed , reducing PHYSICS TESTS provided~ : .-a-nd.. ,. . th_

                                                                                                                        . .,. e_n_u_m...,.b_e_r -of, .---.

the number of required channels ~- THERMAL POWER is

                                                                              ~
                                                                           % RTP; and required channels for LCO 3.3.1, "RTS from 4 to 3.

Instrumentation ,"

b. SDM is within the limits provided in the COLR; and Functions 2, 3, 6, and 17.e, may be reduced

c. RCS lowest loop average temperature is~.

to 3 Farley Units 1 and 2 B 3.1.8-5 Revision~ E2 -15

Joseph M. Farley Nuclear Plant - Units 1 and 2 Response to Request for Additional Information Regarding Multiple Technical Specifications Changes Enclosure 3 Revised Technical Specifications Clean-Typed Pages

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages PHYSICS TESTS Exceptio ns-MODE 2 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptio ns-MODE 2 LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of LCO 3.1.3, "Moderator Temperature Coefficient (MTC)"; LCO 3.1.4, "Rod Group Alignment Limits"; LCO 3.1.5, "Shutdown Bank Insertion Limits"; LCO 3.1.6, "Control Bank Insertion Limits"; and LCO 3.4.2, "RCS Minimum Temperature for Criticality" may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, and 17.e, may be reduced to 3 required channels, provided:

a. THERMAL POWER is s 5% RTP;

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

c. RCS lowest loop average temperature is<:!: 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 hour 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. (Unit 1) Amendment No. (Unit 2) E3 -1 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

 ---------- ---------- -------NO TE------ ---------- ---------- --

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 hour after THERMAL POWER is< P-6. Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2 --------NOTE--------- Not required to be performed until 24 hours after THERMAL POWER is ~ 15% RTP. Compare results of calorimetric heat balance In accordance with calculation to power range channel output. Adjust the Surveillance power range channel output if calorimetric heat Frequency Control balance calculation results exceed power range Program channel output by more than +2% RTP. SR 3.3.1.3 ----------- -----NOT ES------- -------

1. Not required to be performed until 7 days after THERMAL POWER is~ 50% RTP.

2. Performance of SR 3.3.1.9 satisfies this SR.

Compare results of the incore detector In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute difference Frequency Control is~ 3% RTP. Program Farley Units 1 and 2 3.3.1-9 Amendment No. (Unit 1) Amendment No. (Unit 2) E3-2 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Two block valves F.1 Place associated 1 hour inoperable. PORVs in manual control. AND F.2 Restore one block valve 2 hours to OPERABLE status. G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time of Condition F not AND met. G.2 Be in MODE4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ------- --NOT ES

1. Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.

2. Only required to be performed in MODES 1 and 2.

Perform a complete cycle of each block valve. In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.11-3 Amendment No. (Unit 1) Amendment No. (Unit 2) E3-3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages DC Sources- Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.6 -------- -------- NOTE- -------- ----- This Surveillance may be performed in MODE 1, 2, 3, 4, 5, or 6 provided spare or redundant charger(s) placed in service are within surveillance frequency to maintain DC subsystem(s) OPERABLE. Verify each required Auxiliary Building battery In accordance with charger supplies ~ 536 amps at ~ 125 V for~ 4 hours the Surveillance and each required SWIS battery charger supplies Frequency Control

                              ~ 3 amps at~ 125 V for~ 4 hours.                           Program SR 3.8.4.7                ------- ---NO TES-- ------- -----

1. The performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4. 7 once per 60 months.

2. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test at any time.

3. This Surveillance shall not normally be In accordance with performed for the Auxiliary Building batteries in the Surveillance MODE 1, 2, 3, or 4. However, portions of the Frequency Control Surveillance may be performed to reestablish Program OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design load profile described in the Final Safety Analysis Report, Section 8.3.2, by subjecting the battery to a service test. Farley Units 1 and 2 3.8.4-3 Amendment No. (Unit 1) Amendment No. (Unit 2) E3-4

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages DC Sources- Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.8 ------- -----N OTE-- ------- ----- This Surveillance shall not normally be performed for the Auxiliary Building batteries in MODE 1, 2, 3, or 4. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced. Verify battery capacity is ~ 80% of the manufacturer's In accordance with rating when subjected to a performance discharge the Surveillance test or a modified performance discharge test. Frequency Control Program AND 18 months when battery shows degradation or has reached 85% of expected life or 17 years, whichever comes first Farley Units 1 and 2 3.8.4-4 Amendment No. (Unit 1) Amendment No. (Unit 2) E3 -5

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration stated in 10 CFR 20, Appendix B (to paragraphs 20.1001-20.2401),

Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at and beyond the site boundary as follows:

1. For noble gases: Less than or equal to a dose rate of 500 mrem/year to the total body and less than or equal to a dose rate of 3000 mrem/year to the skin, and

2. For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; (continued)

Farley Units 1 and 2 5.5-3 Amendment No. (Unit 1) Amendment No. (Unit 2) E3-6

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABLE not violated. When one or more of the requirements specified in SAFETY ANALYSES LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, (continued) LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to s 5% RTP, the reactor coolant temperature is kept

                                    ~ 531°F, and SDM is within the limits provided in the COLR.

The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. PHYSICS TESTS meet the criteria for inclusion in the Technical Specifications, since the components and process variable LCOs suspended during PHYSICS TESTS meet Criteria 1, 2, and 3 of 10 CFR 50.36 (c)(2)(ii). Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity. LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One Power Range Neutron Flux Channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met. The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6, and 17.e, may be reduced to 3 required channels during the performance of PHYSICS TESTS provided:

a. THERMAL POWER is s 5% RTP; and

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

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

Farley Units 1 and 2 B 3.1.8-5 Revision E3-7

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages RTS Instrumentation B3.3.1 BASES SURVEILLANCE SR 3.3.1.3 REQUIREMENTS (continued) SR 3.3.1.3 compares the incore system to the NIS channel output. If the absolute difference is;;:: 3% RTP the NIS channel is still OPERABLE, but it must be adjusted. The excore NIS channel shall be adjusted if the absolute difference between the in core and excore AFD is t::3% RTP. If the NIS channel cannot be properly adjusted, the channel is declared inoperable. This Surveillance is performed to periodically verify the f(~l) input to the overtemperature ~T Function. Two Notes modify SR 3.3.1.3. Note 1 clarifies that the Surveillance is required only if reactor power is ~ 50% RTP and that 7 days are allowed for performing the Surveillance and channel adjustment, if necessary, after reaching 50% RTP. Apower level of ~ 50% RTP is consistent with the requirements of SR 3.3.1.9. Note 2 allows SR 3.3.1.9 to be performed in lieu of SR 3.3.1.3, since SR 3.3.1.9 calibrates (i.e., requires channel adjustment) the excore channels to the incore channels, it envelopes the performance of SR 3.3.1.3. For each operating cycle, the initial channel normalization is performed under SR 3.3.1.9. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT. This test shall verify OPERABILITY by actuation of the end devices. The RTB test shall include separate verification of the undervoltage trip via the Reactor Protection System and the local manual shunt trip mechanism. The bypass breaker test shall include a local manual shunt trip and local manual undervoltage trip. A Note has been added to indicate that this test must be performed on a bypass breaker prior to placing it in service. The independent test of undervoltage and shunt (continued) Farley Units 1 and 2 B 3.3.1-52 Revision E3-8

Enclosure 3to NL-15-1814 Revised Technical Specifications Clean-Typed Pages Pressurizer PORVs B 3.4.11 BASES ACTIONS D.1 and D.2 (continued) If the Required Action of Condition A, B, or C is not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE. E.1. E.2. E.3. and E.4 If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the Completion Time of 1 hour or isolate the flow path by closing and removing the power to the associated block valves. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time and provides the operator time to correct the situation. If one PORV is restored and one PORV remains inoperable, then the plant will be in Condition B with the time clock started at the original declaration of having two PORVs inoperable. If no PORVs are restored within the Completion Time, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE. F.1 and F.2 If two block valves are inoperable, it is necessary to either restore the block valves within the Completion Time of 1 hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hours. The Completion Times are reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation. (continued) Farley Units 1 and 2 B 3.4.11-6 Revision E3-9

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages Pressurizer PORVs B 3.4.11 BASES ACTIONS G.1 and G.2 (continued) If the Required Actions of Condition F are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6, the PORVs are not required OPERABLE. SURVEr'LLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be closed if needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is modified by two Notes. Note 1 modifies this SR by stating that it is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable. Note 2 modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature conditions, prior to entering MODE 1 or 2. In accordance with Reference 3, administrative controls require this test to be performed in MODE 3 or 4 to adequately simulate opening temperature and pressure effects on PORV operation. (continued) Farley Units 1 and 2 B 3.4.11-7 Revision E3 -10

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages DC Sources- Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.7 (continued) REQUIREMENTS The reason for Note 3 is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance 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 partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or on site system when they are tied together or operated independently for the partial 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 portions of the Surveillance are performed in MODES 1, 2, 3, or 4. Risk insights or deterministic methods may be used for this assessment. SR 3.8.4.8 A battery performance discharge test is a test of constant current capacity of a battery, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage. A battery modified performance discharge test is described in the Bases for SR 3.8.4.7. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8. The modified performance discharge test may be used to satisfy SR 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 at any time. The performance discharge test may be used to satisfy 3.8.4.8 while simultaneously satisfying the requirements of SR 3.8.4.7 once per 60 months. (continued) Farley Units 1 and 2 B 3.8.4-10 Revision E3 -11

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages DC Sources- Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued) REQUIREMENTS The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 9). This reference recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life or 17 years, whichever comes first, the Surveillance Frequency is reduced to 18 months. Degradation is indicated, according to IEEE-450 (Ref. 9), w~en the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is > 10% below the manufacturer's rating. This SR is modified by a Note. The reason for the Note is that performing the Surveillance for the Auxiliary Building batteries would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillan e 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 partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or on site system when they are tied together or operated independently for the partial 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 portions of the Surveillance are performed in MODES 1, 2, 3, or 4. Risk insights or deterministic metho s may be used for this assessment. Farley Units 1 and 2 B 3.8.4-11 Revision E3 -12

Endosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages DC Sources- Operating 8 3.8.4 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6, March 10, 1971.

3. IEEE-308-1971.

4. FSAR, Section 8.3.

5. None.

6. FSAR, Chapter 6.

7. FSAR, Chapter 15.

8. Regulatory Guide 1.93, December 1974.

9. IEEE-450-1980.

10. Regulatory Guide 1.32, February 1972.

Farley Units 1 and 2 B 3.8.4-12 Revision I E3 -13

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages RHR and Coolant Circulation- Low Water Level B 3.9.5 BASES LCO An OPERABLE RHR loop consists of an RHR pump, a heat {continued) exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The LCO requirements are modified by two Notes. The first Note provides an exception to the requirements for one RHR loop to be OPERABLE and one RHR loop to be in operation. This exception is necessary to ensure the RHR System may be realigned as necessary for up to 2 hours to perform the required surveillance testing necessary to verify the RHR System performance in the ECCS injection mode of operation. The second Note permits the RHR pumps to be deenergized for s 15 minutes when switching from one train to another. The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short and the core outlet temperature is limited to > 10 degrees F below saturation temperature. The Note prohibits boron dilution or draining operations when RHR forced flow is stopped. APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System {RCS), and Section 3.5, Emergency Core Cooling Systems {ECCS). RHR loop requirements in MODE 6 with the water level ~ 23 ft are located in LCO 3.9.4, "Residual Heat Removal {RHR) and Coolant Circulation-High Water Level." ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until ~ 23 ft of water level is established above the reactor vessel flange. When the water level is ~ 23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.4, and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions. {continued) Farley Units 1 and 2 B3.9.5-2 Revision E3 -14

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages RHR and Coolant Circulation- Low Water Level B 3.9.5 BASES ACTIONS B.1 (continued) If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur by the addition of water with a lower boron concentration ti;Jan the required boron concentration specified in the COLR. Therefore, actions that could result in the addition of water to the RCS with a boron concentration less than the required boron concentration specified in the COLR must be suspended immediately. If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously. B.3. B.4. B.5.1, and B.5.2 If no RHR is in operation, the following actions must be taken: a) the equipment hatch must be closed and secured with four bolts; b) one door in each air lock must be closed; and c) each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Performing the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded. The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time. (continued) Farley Units 1 and 2 B 3.9.5-3 Revision E3 -15

Enclosure 3 to NL-15-1814 Revised Technical Specifications Clean-Typed Pages RHR and Coolant Circulation- Low Water Level B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. REFERENCES 1. FSAR, Section 5.5.7. Farley Units 1 and 2 B 3.9.5-4 Revision I E3 -16}}

v t e Letter Sequence Response to RAI
CAC:MF5317, (Approved, Closed)
Initiation Acceptance... Administration Questions 30 December 2015 17 May 2016 Answers 28 September 2015 3 March 2016 Results Approval Other:
MONTHYEAR2016-10-20 [Table View]

NL-15-1814, Response to Request for Additional Information Regarding Multiple Technical Specifications Changes

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NL-15-1814, Response to Request for Additional Information Regarding Multiple Technical Specifications Changes

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