View source for L-99-021, Forwards Response to RAI Re Conversion to ITSs for Chapter 3.3.Attachment II Includes Proposed Revs to Previously Submitted LAR Re Rais,Grouped by RAI number.Clean-typed Copies of Affected ITS Pages Not Included

{{Adams
| number = ML20207D610
| issue date = 05/28/1999
| title = Forwards Response to RAI Re Conversion to ITSs for Chapter 3.3.Attachment II Includes Proposed Revs to Previously Submitted LAR Re Rais,Grouped by RAI number.Clean-typed Copies of Affected ITS Pages Not Included
| author name = Morey D
| author affiliation = SOUTHERN NUCLEAR OPERATING CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000348, 05000364
| license number = 
| contact person = 
| document report number = NEL-99-0212, NEL-99-212, NUDOCS 9906040027
| package number = ML20207D618
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 32
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:Dave Morey                 Southern Nuclear Vice President             Operating Company. Inc.
f arley Project            Post Othce Box 1295 Bnmingham Alanama 352W
   .                                      Tel 205 992 5131 May 28,1999                                                                    SOUTHERN L                  k          %
COMPANY                       _
Energy to Sers e You r World '
* i Docket Nos. 50-348                                                                      NEL-99-0212 4
50-364 U. S. Nuclear Regulatory Commission                                                                                                 _
ATTN.: Document Control Desk                                                                                                        -
Washington, DC 20555-0001                                                                                              7 Joseph M. Farley Nuclear Plant                                              #
Response to Request for Additional Information Related to Conversion te 'he Imoroved Technical Specifications - Chanter 3.3 k
7 Ladies and Gentlemen:                                                                                                  I By letters dated March 12,1998, and April 24,1998, Southern Nuclear Operating Company                                           ,
                                                                                                                                            ~
(SNC) submitted the Farley Nuclear Plant (FNP) - specific Improved Technical Specifications (ITS) conversion documentation packages in accordance with 10 CFR 50.90. The {{letter dated|date=April 24, 1998|text=April 24,1998 letter}}, which s' bmitted the Clean-Typed copies of the FNP iTS, included an attachment which provided hard copies of changes to the original submittal to correct minor editorial errors and inconsistencies within the package. By {{letter dated|date=August 20, 1998|text=letter dated August 20,1998}}, SNC submitted an                                   4 electronic copy of the Discussion of Changes (DOCS) and Significant Hazards Evaluations                                         -
(SHEs) associated with the ITS conversion. Included with that letter were hard copies of changes                       7 to the original submittal to correct additional minor editorial errors and inconsistencies within the                      -
package. By {{letter dated|date=November 20, 1998|text=letter dated November 20,1998}}, SNC submitted responses to a Request for                                       =
Additional Information (RAI) for Chapters 3.6 and 5.0. Included with that letter were hard copies of changes to the original submittal to reflect the SNC responses to the RAI. By {{letter dated|date=February 20, 1999|text=letter dated February 20,1999}}, SNC submitted responses to a RAI for Chapter 3.4. Included with that letter were hard copies of changes to the original submittal to reflect the SNC responses to the RAI. By letters (2) dated April 30,1999, SNC submitted responses to RAls for Chapters 3.1,3.2,3.5,3.7, 3.8, and 3.9. Included with those letters were hard copies of changes to the original submittal to reflect the SNC responses to the RAls. The NRC StafTrcquested that SNC provide additional f/.
                                                                                                                                 /
information for Chapter 3.3 via E-mail dated March 3,1999.                                                          /        -
Attachment I provides the SNC responses to the NRC RAI questions for Chapter 3.3.                                        -
Attachment II includes proposed revisions to the previous!y submitted license amendment request                          -
related to these RAls, grouped by RAI number. During meetings held on April 19-20,1999, the                     g(M                -
Stafistated that it was not necessary to provide mark-ups of the Current Technical Specifications               I (CTS) in responses to RAls. Therefore, the attached pages do not contain CTS mark-ups.                                  4 Attachment III provides changes made to the submittal for Chapter 3.3 to address SNC identified                          -2 issues, editorial changes, omissions, and inconsistencies in the package identified after the original submittal.
9906040027 990528 PDR      ADOCK 050003 8 m

w-                                                                                                      ;
l l
Page 2 U. S. Nuclear Regulatory Commission In response to this RAI, some changes to the SHEs were required. SNC has determined the proposed changes to the FNP TS do not involve a significant hazards consideration as defined by 10 CFR 50.92(c). The revised SHEs are included in Attachment II. SNC has also determined that the proposed changes will not significantly affect the quality of the human environment. A       1 copy of the proposed changes ha; been sent to Dr. D. E. Williamson, the Alabama State Designee, in accordance with 10 CFR 50.91(b)(i).
Clean-typed copies of the affected ITS pages are not included. A complete clean-typed copy of the FNP ITS will be re-submitted at the end of the NRC review process.
Mr. D. N. Morey states that he is a Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company and that, to the best of his knowledge and belief, the facts set forth in this letter and attachments are true.
If there are any questions, please advise.
l_ Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY hlV Dave Morey Skorn to andsubscribedbefor me thi               >
999 Y/d
       'K 0, N V
NotaryPublic My Commission Expires: Y s
: l. N 0 f WAS/maf:ITSRAI,_5. DOC Attachments I.      SNC Response to NRC Request for Additional Information Related to Conversion to the Improved Technical Specifications - Chapter 3.3
            - II. SNC Response to NRC Request for Additional Information Related to Conversion to the Improved Technical Specifications, Chapter 3.3 - Associated Package Changes Grouped bv RAI Number                                               .
III. SNC Identified Changes - Associated Package Changes l
4 cc:    See next page.
I 1
h                                                                                                     J

cc: Southern Nuclear Ooeratina Cornoany Mr. L. M. Stinson, General Manager - Farley _
U. S. Nuclear Regulatory Commission. Washington. D. C.
Mr. L. Mark Padovan, Licensing Project Manager - Farley U. S. Nuclear Regulatory Commission. Renion II Mr.L. A.Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D, E. Williamson, State Health Officer l
1 1
4 I
1 l
i

ATTACHMENT I SNC Response to NRC Request for AdditionalInformation Related to Conversion to the Improved Technical Specifications - Chapter 3.3 J

r' SNC Respinse to NRC RAI Related ts Larpter 3.3 NRC Question No. 3.3.1-1 1 

==Reference:==
ITS Table 3.3.1 1 Notes 1 and 2 CTS Table 2.2-1 Notes 1 and 2 DOC L.9 and JFD 25 CTS Table 2.2-1 Notes 1 and 2 provide equations for Overtemperature and Overpower AT trip setpoints. He corresponding ITS Table 3.3.1 1 Notes 1 and 2 provide Overtemperature and-Overpower AT trip setpoint calculations. Several changes have been made in the conversion              I from CTS Table 2.2-1 Notes 1 and 2 to ITS Table 3.3.1-1 Notes I and 2 without adequate discussion andjustification. He revision consists of changing some of the terms used in these formulas from an equality to an inequality. Dese changes also result in deviations from the STS.
DOC L.9 and JFD 25 provide discussion andjustification for these changes and state that the           3 changes consider the applicable safety analyses. It is assumed that the CTS formulas are consistent with assumptions of the FNP safety analyses, and that changes to the CTS formulas affect these assumptions. Example changes are:
M                             IIS                         3.T.S P' = 2235 psig                P' 2 2235 psig              P's 2235 psig K 3= 0.000825                 K 32 0.000825               K 3= 0.000671 Examples of other terms affected by this change include Ki, K2, K5 Ks,and L.
                                                                                   ~
Comment:        Provide additional discussion and justification for the changes in equality / inequality of the setpoint terms.
SNC Response No. 3.3.1-1 Based on this NRC review comment, SNC evaluated the initial FNP ITS submittal including DOC 9-L and JFD-25. SNC determined that the Overtemperature AT and Overpower AT (OTAT
       & OPAT) equations must be expressed with a preference for the current licensing basis and CTS with a secondary consideration for upgrading to STS format. As such, the ITS and STS Reactor           i Trip System (RTS) OTAT & OPAT setpoint equations described in Notes 1 and 2, the associated Bases, significant hazards evaluation, DOC 9-L, and JFD-25 have been revised. To conform with the current Farley design basis, equalities will be maintained for Ki - K6 and P'; inequalities will be maintained for 1,  5 26 , T, and T'; and the equalities for t ,i v2. and t3will be changed to inequalities. A discussion of the STS deviations and the proposed change is presented herein.
STS Table 3.3.1-1 Notes I and 2, which provide the detailed OTAT & OPAT setpoint equations,            ,
I are revised based on FNP design basis. He equation revisions, including the defined terms, static and dynamic constants and associated values, reflect Farley-specific hardware, accident analyses modeling and assumptions, setpoint uncertainty calculations, scaling calculations, and calibration     ,
and periodic surveillance practices. The Farley-specific approach provides assurance that the         l OTAT & OPAT reactor trip functions will respond conservatively with respect to the safety             i analyses that credit these trip functions. He Faricy-specific scaling and calibration methods ensure that these RTS functions are set conservatively with respect to the setpoint equation inequality provided in the Technical Specifications. Where appropriate, the equation and term         i descriptions will be revised in the Farley ITS to conform more closely to the STS format. De          !
technical basis for the Farley STS deviations and time constant equality / inequality changes follow.
Page 1 of 22 J

SNC Resp:nse ts NRC RAI Rel:t:d 13 Ch:pt:r 3.3 The OTAT & OPAT dynamic compensation term STS deviations reflect the as-built Farley hardware configuration as approved in Technical Specifications Amendment Nos. 87 (Unit 1) and 85 (Unit 2), FNP RTD Bypass Manifold Elimination.
The equalities, inequalities and vaiues associated with the static and dynamic constants and allowable values were approved in Technical Specifications Amendment Nos.121 (Unit 1) and 113 (Unit 2), Revision to Core Limits and OTAT & OPAT Setpoints.
The use of equalities for the static constants Ki - K6 is required because these are explicit scaling calculation inputs that must be implemented within the hardware calibration tolerances. The supporting setpoint uncertainty calculations reflect the static constant gain values and the hardware calibration tolerances and methods, and demonstrate margin to the corresponding OTAT & OPAT safety analysis limits presented in FSAR Tab!c 15.1-1 A.
He change to the dynamic time constants ti , ta and t 3will ensure that these time constants are conservatively calibrated with respect to the safety analyses modeling. This approach is              I I
necessary because the dynamic time constants are modeled as explicit inputs in the safety analyses. As such, the current scaling calculations and calibration procedures specify a time constant setting with a calibration tolerance on the conservative side of the inequality. Therefore, the Technical Specifications should specify the conservative direction for the calibration of dynamic time constants ti.1.2 and t .3 (Note that previous changes to t 3and 16 time constant values included inequalities; refer to Amendment Nos.121 (Unit 1) and 113 (Unit 2).)
The requirement to set T' and T" to equal the full power operating reference temperature (T r) was also approved in these amendments, along with the clarification that ATomust be normalized at the full power operating reference temperature. This approach is required to ensure that the temperature correction factors (i.e., setpoint penalties for changes in specific heat variations from full power calibration conditions) are set to provide accurate correction factors consistent with accident analysis assumptions.
The inequality requirement to set T' and T" 5.577.2 'F is retained to ensure that Farley is not operated beyond its analyzed steady-state full power reference temperature.
In addition to the above changes, other required changes were identified during this review. They are as follows:                                                                                       j The term descriptions for AT, T, and ATo are modified to clarify that these terms for each RTS protection channel are associated with specific RCS " loops," versus a "RCS" average AT or            ,
average Tm.                                                                                           l Tue term description for P' is modified to clarify that the actual process control parameter is the
  " pressurizer" pressure reference setpoint (i.e., Por). He use of an equality for the nominal operating pressurizer pressure value reflects the transient and accident analyses modeling of a nominal setpoint with a plus/minus uncertainty. Plant procedures implement the pre;sure               ;
reference setpoint within required calibration tolerances, and the control system setpoint uncertainty calculations demonstrate that the accident analyses pressure control uncertainty assumption is conservative. (Refer to WCAP-12771, Revision 1," Westinghouse Revised Hermal Design Procedure Instrument Uncertainty Methodology for Alabama Power Farley Nuclear Plant Units 1 and 2 (Uprating to 2785 MWT NSSS Power).")
Page 2 of 22

SNC Response 12 NRC RAI Rel:ted to Ch'pt:r 3.3 Retention of the f i (AI) term description clarifies that it is an " indicated" function which must be calibrated (i.e., normalized) based on cycle-specific test data following each refueling. Plant procedures ensure that fi (AI) is properly calibrated based on the incore/excore cross-calibration data. The OTAT & OPAT setpoint uncertainty calculations explicitly account for the cycle-specific f (AI) i   cal!* oration process.
These Farley-specific deviations from STS and the dynamic time constant equality / inequality changes provide assurance that FNP is operated and the OTAT & OPAT reactor trip functions are maintained consistent with applicable FNP transient and accident analyses, safety analysis limits, and setpoint uncertainty and scaling calculations. Hese STS deviations and the dynamic time constant changes also reflect the current FNP design basis.
NRC Question No. 3.3.1-2

==Reference:==
ITS 3.3.1 Required Actions D.3 and E.2 CTS Table 3.3-1 Actions 2c or 7a                                                      ,
DOC 62-M and 66a-M Failure to perform CTS Table 3.3-1 Actions 2c or 7a, would require entry into CTS 3.0.3, allowing 6 hours to be in MODE 3. He corresponding ITS 3.3.1 Required Actions, D.3 and E.2 allow 12 hours to be in MODE 3 under the same conditions. His extension of Completion Time is discussed in DOC 62-M as a more restrictive change. His additional 6 hours for entry into MODE 3 results in a Less Restrictive Change, that is consistent with the STS.
Comment:         Provide appropriate Less Restrictive Change documentation.
SNC Response No. 3.3.1-2 SNC believes that the addition of new LCO 3.3.1 default Actions D.3 and E.2 is a more restrictive ITS change. CTS Table 3.3-1 Actions 2 and 7 do not have an explicit action statement for the condition when Action 2 or 7 is not satisfied. Given such a condition, the CTS would
   - require entry into 3.0.3. He limiting CTS Table 3.3-1 Action 2 and 7 time is 6 hours to place the affected channel in trip (Actions 2c and 7a). If the action is not completed, then LCO 3.0.3 is entered. The 3.0.3 Action requires initiation of power reduction within 1 bour and, subsequently, placing the Unit in Mode 3 within the next 6 hours. As such, the CTS total allowed time to achieve Mode 3 is 13 hours; whereas, the STS default allowance is 12 hours. Herefore, the major impact of the proposed addition to the Farley ITS is seen as a more restrictive change to the CTS. Justifications DOC No. 62-M and DOC No. 66a-M have been enhanced to reflect the above discussion.
As discussed in the April 19,1999 meeting with the Staff, not entering LCO 3.0.3 eliminates the associated reporting requirement of 10 CFR 50.73. With respect to the reduction in regulatory burden due to the elimination of this requirement, a generic DOC, type "LC" has been created and is included, along with the associated SHE, in Attachment II.
NRC Question No. 3.3.1-3

==Reference:==
ITS Table 3.3.1-1 Conditions O and P CTS Table 3.3-1 Actions 6 and 7 DOC 64-M Page 3 of 22 i.

=
SNC Resp =se ts NRC RAI Reirt:d ta Ch:pter 3.3 CTS Table 3.3-1 Actions 6 and 7 require placing an inoperable Turbine Trip instrument channel in trip within 6 hours, or entry into CTS 3.0.3 as the default Action is required.. De corresponding ITS Table 3.3.1-1 Conditions O and P maintain the requirement to trip the channel within 6 hours but provide an alternative default Action, to reduce power to less than the P-9 setpoint. His change, discussed in DOC 64-M as a More Restrictive Change is actually a less Restrictive Change, because an alternative to shutdown per 3.0.3 is allowed. In addition, the 10-hour Completion Time for reducing power to less than P-9 is not discuned orjustified.
Comment: Provide appropriate documentation for the Less Restrictive Changes.
SNC Response No. 3.3.1-3 SNC concurs with the Staffs assessment. De STS turbine trip / reactor trip function Required Action is to reduce power below P-9 if an inoperable channel is not placed in trip within 6 hours.
l  The Farley CTS, as per LCO 3.0.3, require power reduction and placing the Unit in Mode 3 if the inoperable channel is not placed in trip within 6 hours. Therefore, the proposed ITS change is a less restrictive change.
L  DOC NOs. 37-L and 64-M have been revised to clarify that the STS default is a less restrictive change. DOC NO. 37-L provides the basis for the new FNP default action.
i l* With regard to the Completion Time for Required Actions O.2 and P.2, SNC believes that the l  new default action times are a more restrictive ITS change. Also, the proposed ITS Completion l  Time for these Actions is identical to the STS. CTS Table 3.3-1 Actions 6 and 7 do not have an explicit action statement for the condition where Action 6 or 7 is not satis 6ed. Given such a condition, the CTS would require entry into 3.0.3. De limiting CTS Table 3.3-1 Action 6 and 7 time is 6 hours to place the affected channel in trip (Actions 6 and 7a). If the Action is not l
completed, then LCO 3.0.3 is entered. The 3.0.3 Action requires initiation of power reduction
!- within I hour and, subsequently, placing the Unit in Mode 3 within the next 6 hours. De STS required time allows 6 hours to place the channel in trip and an additional 4 hours (based on        )
l  operating experience) to reduce power to a level (< =50% RTP) where the turbine trip / reactor trip function is not applicable. As such, the CTS default is a total allowed time of 13 hours; whereas,   l the S'I3 default allowance is 10 hours. Since both default Actions require that power be reduced to a Condition where this trip function is not required and the STS Action time is more limiting,    l the proposed inclusion of the STS 10 hour Co:npletion Time for the new default Action is more        ;
restrictive.
DCC NO. 64-M has been revised to reficct the above discussion and to clarify that DOC NO. 64-        ,
M is applicable to both the turbine auto stop oil pressure and the turbine trip throttle RTS         i functions.
NRC Question No. 3.3.1-4

==Reference:==
ITS Table 3.3.1-1 Action M.2 CTS Table 3.3-1 Action 11 DOC 71-M                                                                            I CTS Table 3.3-1 Action 11 does not state a default action; therefore, entry into CTS 3.0.3 would be required. CTS 3.0.3 requires shutdown to MODE 3 within 7 hours. De corresponding ITS Table 3.3.1-1 Action M.2 provides an alternative default action to reduce power to less than P-7 within the following 6 hours. His change discussed as a More Restrictive Change in DOC 71-M allows an alternative to CTS 3.0.3 shutdown, resulting in a Less Restrictive Change.
Page 4 of 22 9

I SNC Respuse't3 NRC RAI Relited 13 Chrpt:r 3.3
    - Comment:         Provide appropriate docamentation for the Less Restrictive Changes.
SNC Response No. 3.3.1-4 SNC concurs with the Staffs assessment. The STS reactor coolant pump breaker position reactor
: trip function Required Action is to reduce power below P-7 if an inoperable channel is not placed in trip within 6 hours. Table 3.3-1 Action 11 of the Farley CTS, as per LCO 3.0.3, requires power reduction and placing the Unit in Mode 3 if the inoperable channel is not placed in trip within 6 hours. 'Iherefore, the proposed ITS change is a less restrictive change.
As indicated on the CTS markups, DOC NO. 41-L provides the basis for the new FNP default action. DOC NO. 71-M has been revised to reference this basis. In addition, DOC NO. 71-M has been revised to provide the technicaljustification for a more restrictive Action time duration With regard to the Completion Time for Required Action M.2, SNC believes that the new default action time is a more restrictive ITS change. Also, the proposed ITS Completion Time is identical to the STS. The limiting CTS Table 3.3-1 Action 11 time is 6 hours to place the affected channel in trip. If the Action is not completed, then LCO 3.0.3 is entered. The 3.0.3 Action requires initiation of power reduction within I hour and, subsequently, placing the Unit in Mode 3 within the next 6 hours. The STS required time allows 6 hours to place the channel in trip and an additional 6 hours (based on operating experience, redundant capability, and low event occurrence probability) to reduce power to a level (< =10% RTP) where the reactor coolant pump breaker position reactor trip function is not applicable. As such, the CTS default is a total allowed time of 13 hours; whereas, the corresponding STS default allowance is 12 hours. Since both default Actions require that power be reduced to a Condition where this trip function is not
    ' required and the STS Action time is more limiting, the proposed inclusion of the STS 12 hour Completion Time for the new default Action is more restrictive.
Justification DOC NO. 71-M has been revised to reflect the above discussion.
: With respect to the reduction in regulatory burden due to the elimination of the requirement to
    . report an entry into LCO 3.0.3, a generic DOC, type "LC" has been created and is included, along with the associated SHE, in Attachment II (see markups associated with SNC Response No.                !
3.3.1-2).                                                                                              j NRC Question No. 3.3.15

==Reference:==
ITS 3.3.1 Condition R STS 3.3.1 Condition R Note 2 CTS Table 3.3-1 Action [14]
JFD-7                                                                                 ;
i CTS Table 3.3-1 Action [14] provides a requirement that a RTB shall not be bypassed while one of the diverse features is inoperable, except for the time required to perform maintenance to restore the breaker to OPERABLE status. This requirement is used in JFD-7 tojustify bypassing
: a breaker for maintenance with no time limitation. STS 3.3.1 Condition R Note 2 limits bypassing a breaker for up to 2 hours in this condition. .'Ihis 2 hour time restriction is deleted in ITS 3.3.1 Condition R. 'Ihis change, resulting in a deviation from the STS is considered unacceptable without plant specificjustification. It is not clear that intent of the CTS Action [14]
wording is to allow an unlimited time to remove a breaker from [ service using] bypass, and the        ;
deviation from the STS is not adequatelyjustified.
                                                 ' Page 5 of 22 j

SNC R:sp::se to NRC RAI Rtirt:d ta Chrpttr 3.3
  . Comment: Provide additional discussion and justification for the CTS change based on Safety Analysis assumptions for RTB Operability.
SNC Response No. 3.3.1-5 Based on further reviews, SNC has determined that STS Condition R Note 2 must be revised and relocated to provide assurance that the current Farley licensing basis, as reflected in the last sentence of CTS Table 3.3-1 Action 14, is maintained consistent with the Farley design. In addition, the associated ITS Bases have been revised to reflect the Farley-specific licensing basis.
As a result, the associated ITS pages have been resised, along with supporting justifications DOC NO. 75A, JFD NO. 7 and JFD NO. 7b.                                                                   j STS Condition R, which applies to the Reactor Trip Breaker (RTB) Trains, contains two notes.
Note 2 allows a given RTB to be bypassed for maintenance on an inoperable undervoltage or shunt trip mechanism for up to 2 hours if the other RTB train is operable. The proposed ITS deviations from STS revise and relocate Note 2 to be consistent with the currer.t Farley design and licensing basis as reflected in the last sentence of CTS Table 3.3-1 Action 14.
i First, Note 2 is revised by deleting the explicit time duration for bypass because CTS Table 3.3-1   l Action 14 includes no specific time duration. The unrestricted bypass allowance is provided to give sufEcient time to accomplish corrective maintenance. However, the expectation is that such      ;
maintenance would be accomplished in a timely manner, and the explicit CTS requirement is that affected RTB would only be bypassed during actual maintenance activities. The technical basis for this allowance is found in NRC Generic Letter 85-09," Technical Specifications For Generic Letter 83-28, Item 4.3." Based on the current licensing basis as approved in Tecimical               l Specifications Amendment Nos. 67 (Ul) and 59 (U2), the exclusion of the 2 hour bypass                i allowance is acceptable.
Second,' minor editorial changes are incorporated in STS Condition R Note 2. Rese changes            l I
clarify that the bypass allowance applies only to one of two diverse trip mechanisms associated with a given RTB when the other RTB Train is Operable.
Third, STS Condition R Note 2 is relocated to ITS Condition U based on the following. STS 3   Condition R pertains to the RTB Trains. However, Note 2 is directly applicable to the undervoltage and shunt trip mechanisms, which are the portions of the RTB Trains excluded from       ;
STS Function No.18. He diverse trip mechanisms are listed as STS Function No.19, which utilizes STS Condition U. Because Farley listed the undervoltage and shunt trip mechanisms as a new ITS Functional Unit to conform to STS and CTS Action 14 applies directly to these diverse
  ' trip mechanisms, Note 2 must be relocated to ITS Condition U.
These ITS deviations from the STS assure that the Farley licensing basis for CTS Table 3.3-1 Action 14 is retained consistent with the Farley design.
NRC Question No. 3.3.1-6

==Reference:==
ITS 3.3.1 Condition V JFD-24
   . ITS 3.3.1 Condition V is added to address two inoperable RTS trains, including RTBs and Automatic Trip logic. This change discussed in JFD-24 results in a deviation from the STS.
Page 6 of 22

y .- - - ,    . .     .
t
   ,                                         : SNC Resporse is NRC RAI Rel:ted to Chapter 3.3 JFD-24 identifies this as an error in the STS. This change [is] not discussed orjustified on plant specific design basis.
Comment: Provide additional discussion and justification for the change based on plant specific design basis.
SNC Response No. 3.3.1                       De discussion provided in JFD NO. 24 states that Condition V was incorporated into the RTS Condition subsection in STS Revision 1 based on an NRC change notice (NRC-02 01) and that it .
was inadvertently omitted from the Conditions column for Function Nos.19 and 21 in RTS Table 3.3.1-1. Condition V was also added to the STS Bases in Revision 1, which further substantiates the basis for applying Condition V to the RTS Reactor Trip Breaker and Automatic Trip Logic
                    < Trains. DOC NO. 47-M states that Condition V is added to the ITS to be consistent with the STS L and its guiding principles.- DOC NO. 47-M also indicates that Condition V is applicable to the Farley RTS and that the inclusion of Condition V will provide clarification re8arding required Action when multiple Conditiou entries are encounter. Therefore, RTS Condition V was incorpois,;cd into the Farley ITS exactly as intended by STS Revision 1. SNC believes that the information presented in JFD NO. 24 and DOC NO. 47-M provide sufficient justification and no changes to thejustification are required. However, the applicability to Condition V should also apply to Modes 3,4, and 5. Derefore, ITS Table 3.3.1-1 has been revised to add a reference to Condition V for ITS Functions 18 and 20 in these applicable Modes.
NRC Question No. 3.3.1-7

==Reference:==
STS 3.3.1 Action U.2.2 JFD-8 STS 3.3.1 Required U.2.2 requires RTBs to be opened one hour aAer shutdown to MODE 3 per Required Action U.2.1. This requirement to open the RTBs is deleted in the ITS. This deviation from the STS is not discussed andjustified in JFD-8 on a plant specific basis.
Comment: Provide additionaljustification for the change based on plant specific design basis.
                   . SNC Response No. 3.3.1-7 As ' documented in the justifications provided by JFD 8, DOC NOS. 46-A, 53-L,54-A, 55-M and 73-A, and the corresponding ITS Bases, this deviation from STS is required to maintain consistency with the current Farley licensing basis and CTS. In addition, Condition U applies to
                   - Function 18, Reactor Trip Breakers, in MODES 1 and 2. Once MODE 3 is entered, the modes of applicability for this Condition are exited. In MODE 3, Condition C applies which contains an Action to open the RTBs if the channel or train is not returned to operable status within the associated Completion Time. The following describes the basis for the change.
De Farley ITS deviation eliminates Action U.2.2 from STS Table 3.3.1-1, ne discussion provided in JFD NO. 8 states that the revision to STS Condition U (i.e., Required Action U.2.2)
                   - maintams consistency with CTS Table 3.3-1. The discussion also states that STS Condition U, with the exception of Action U.2.2, is equivalent to CTS Action 14 and that STS Condition C is equivalent to CTS Action 13. DOC NO. 46-A confirms the applicability of STS Condition U to the Farley design and CTS Action 14. DOC NO. 73-A, which has been revised to clarify that 0    . CTS Action 13 is also applicable to the RTBs and the diverse trip mechanisms, confirms the applicability of STS Condition C to the Farley design and CTS Action l'8.- DOC NOS. 53-L,54-Page 7 of 22 L-                                                    _ _ _ . _ _ _ _

n
                                         ' SNC R:.sponse ts NRC RAI Relit:d to Chrpter 3.3 A and 55-M confirm the applicability of STS Condition R to the Farley design and CTS Action 1.
De JFD NO. 8 discussion explains how the STS Condition for the RTB trip mechanisms for Modes 1 and 2 (Condition U) and Modes 3,4 and 5 (Condition C) work together in the same manner as the CTS Actions for Modes 1 and 2 (Actions 1 and 14) and for Modes 3,4 and 5
               . (Action 13). JFD NO. 8 has been expanded to include the fact that once Mode 3 is entered, the 3            applicable Modes (Modes 1 and 2) for Condition U have been exited.
A minor revision has been incorporated into DOC NO. 73-A. He revision clarifies that STS Condition C is applicable to the trip breakers and the diverse trip mechanisms, as well as the manual actuation and automatic trip logic. His clari6 cation is appropriate in that CTS Action 13 is applicable to these RTS functions, as shown in CTS Table 3.3-1 and as described in ITS Bases C.1 and C.2, which conform exactly to the corresponding STS Bases.
NRC Question No. 3.3.1-8 '

==Reference:==
 - ITS SR 3.3.1.2 STS SR 3.3.1.2 CTS Table 4.3-1 footnote (2)
CTS Table 4.31 footnote (2) requires a heat balance comparison above 15% RTP for the Power Range instrument channels. No time requirement is included in the CTS aRer exceeding 15%
RTP. The corresponding STS SR 3.3.1.2 requires this comparison but specifies a time limit of 12 hours aRer exceeding 15%. ITS SR 3.3.1.2 increases the STS limit of12 hours to 24 hours.
His STS deviation is considered unacceptable without plant specificjusti6 cation for the change.
              ' Comment:           Provide additionaljustification based on plant specific design.
SNC Response No. 3.3.1-8 DOC NO. I10-M and JFD-9 providejustification for this deviation from the STS and identify the deviation as a more restrictive change. The Farley CTS require performance of a secondary heat balance (i.e., calor metric) and, if necessary, NIS Power Range (PR) channel adjustment (i.e.,
calibration / normalization) on a 24 hour basis when greater than 15% RTP. However, there is no explicit time duration (limit) included with the power level limit. For the Farley ITS, inclusion of a time limit is an allowance that provides relief from immediate compliance with the Technical
: Specifications NIS PR daily surveillance during power ascension. De proposed 24-hour time limit is a reasonable restriction that ensures timely performance of the NIS PR channel calibration, and the proposed time period is consistent with the daily surveillance requirement.
l
            ,    For a given Farley unit stanup at BOL, the initial NIS PR calibration is performed during a power ascension hold at about 30% RTP. This practice and the allowance of 24 hours to achieve this power level are acceptable based on the following. Calorimetric' calculation results at low power levels can include large uncertainties; therefore, it is desirable to raise power to higher levels
: where the calorimetric calculations will be more accurate (Refer to Westinghouse Technical            .i Bulletin ESBU-TB-92-14-RI,"Decalibration Effcets of Calorimetric Power Measurements on the               l NIS High Power Reactor Trip at Power Levels Less than 70% RTP"). In addition to the
               ' calorimetric uncertainty concerns, there are also potential calibration uncertainty effects attributed to the core design. Derefore, plant procedures provide preliminary BOL PR calibration data for initial startup based on cycle-specific core designs and test data. Independent methodology
: review con 6rms, and operating experience demonstrates, that this practice is reliable.
Performance of the initial NIS calibration must also be coordinated with other unit startup -
activities such as plant % y stabilization, NIS incore/excore cross calibration, NIS                     j Page 8 of 22                                               j i                                                                                                                       !
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i SNC Response 13 NRC RAI Rel:ted 13 Chapter 3.3 Intermediate Range calibration, and feedwater control system transitions. Furthermore, the plan       I power ascension rate is limited when raising power above 20% RTP to ensure proper fuel preconditioning. Based on these Farley-specific controls, the proposed 24-hour time allowance is reasonable and necessary to perform this surveillance, in conjunction with other startup activities, without placing the plant in an unsafe condition.
j JFD-9 has been revised to include a summary of the more-detailed Farley-specific justification presented above. In addition, the Bases for SR 3.3.1.2 is revised to indicate that power distribution changes are the primary effect for day-to-day measurements as opposed to instrument drift.
i NRC Question No. 3.3.1-9                                                                               )
l

==Reference:==
ITS SR 3.3.1.9                                                                              l STS SR 3.3.1.6                                                                         j JFD 12a and JFD 11                                                                     l STS SR 3.3.1.6 requires a calibation of excore channels to agree with incore detector measurements every 92 days. "Ihe STS SR 3.3.1.6 Note provides an allowance that the surveillance is not required until 24 hours after THERMAL POWER is greater than or equal to 50% RTP. This surveillance is renumbered in the ITS to ITS SR 3.3.1.9. ITS SR 3.3.1.9 extends the STS surveillance interval of"92 days" to "18 months" (Reference JFD 12a) and extends the 24 hour allowance to 7 days (Reference JFD 11). There is no direct requirement in the CTS to           l perform this surveillance. This surveillance, added to conform to the STS, should maintain the        I surveillance frequency and 24 hour allowance stated in the STS SR 3.3.1.6 Note.                        j Comment: Provide additional technicaljustification for change in surveillance frequencies (24 hours and 92 days).
SNC Response No. 3.3.1-9 Based on further review of the proposed changes associated with ITS surveillance SR 3.3.1.9, SNC has determined that the proposed surveillance frequency (18 months) and allowance (7 days) are appropriate for FNP. In addition, SNC determined that further changes are required to ITS surveillance SR 3.3.1.3 and the ITS Bases for surveillances SR 3.3.1.3, SR 3.3.1.9 and SR 3.3.1.10. The STS and Bases mark-ups have been revised to incorporate these additional changes. In addition, thejustifications provided in DOC NOS. 84-A,86a-L,and 88a-M and JFD NOS.10,11, and 12a have been revised. The following response addresses the above comments.
The CTS do not include a surveillance requirement equivalent to STS SR 3.3.1.6. Nevertheless, Farley routinely performs an incore/excore cross-calibration at BOL for each operating cycle to compensate for changes in core design and fuel assembly manufacturing that affect core power distributions. The cross-calibration ensures that the NIS power range channels AI indications and the f(AI) inputs to the OTAT reactor trip are calibrated based on the cycle-specific core power distributions. Because of fuel operational limitations, the test must be performed at part-power (i.e., approximately 35% to 50% RTP). To assure validity of the calibration data for each NIS power range upper and lower detector output, the excore channel AI calibration is based on analysis of multiple incore flux maps over a range of power distributions. The FNP digital power range meter design also allows for improved incore/excore cross calibration accuracy. 'Ihe technical basis for the Farley incore/excore cross-calibration method was provided to the NRC Staffin SNC letter, " Joseph M. Farley Nuclear Plant Reply to Inspector Follow-up Items Report Number 50-348,364/92-14 Item Numbers 50-348,364/92-14-02 and 92-14-03," dated July 29, Page 9 0f 22 i

7 SNC Resp:nse to NRC RAI Related to Chapter 3.3 1992 [The NRC Staff accepted the basis for the Farley cross-calibration in letter, "NRC Inspection Report Nos.'50-348/94-19 and 50-364/94-19," dated August 25,1994]. De accuracy of tie incore/excore calibration is reflected as an explicit process measurement allowance in the Farley-specific OTAT setpoint uncertainty calculation. This calculation was included in SNC licensing submittal, " Joseph M. Farley Nuclear Plant Technical Specifications Change Request i Revision to Core Limits and OTAT & OPAT Setpoints and Implementation of RAOC," dated June 12,1996 [He NRC Staff approved the requested changes in Amendment Nos.121 (Ul) and 113 (U2)].
To conform more closely to the STS, FNP will include STS SR 3.3.1.6 as ITS SR 3.3.1.9 with a surveillance frequency of 18 months. Normally this incore/excore calibration surveillance will be performed at BOL to normalize the excore channel Al indications and f(AI) inputs for the OTAT reactor trip by adjusting the excore channels to match the cycle-specific core power distributions.
The " bracketed" frequency of 92 EFPD in proposed ITS surveillance SR 3.3.1.9 is changed to 18 months. His frequency is consistent with the current Farley practice. Operating experience at FNP has proven this 18-month frequency to be adequate for performing the incore/excore cross-calibration and for establishing the BOL cycle-specific power range channel Al calibration. He Farley monthly calibration checks and re-normalization, if necessary, of the power range Al channels, required by ITS SR 3.3.1.3, address the effects of flux re-distribution with burnup.
Furthermore, the Farley-specific setpoint calculations include an explicit process measurement allowance of 3% difference between the incore and excore instruments. His allowance is consistent with periodic verification required by SR 3.3.1.3. In addition, performance of the periodic calibrations at a frequency of 92 EFPD would require that Farley operate once per quarter below full power for about 3 days to obtain the extensive incore flux map data required for a multiple-point cross-calibration. Herefore, a frequency of 18 months is appropriate for Farley. His frequency maintains consistency with the current Farley practices and procedures and the supporting setpoint uncertainty calculations.
De " bracketed" time of 24 hours in proposed ITS surveillance SR 3.3.1.9 is changed to 7 days.
This time is reasonable based on the actual time required to: perform multiple flux maps over a range of power distributions at part-power; analyze the resultant data and calculate new excore detector calibration currents; revise NIS power range calibration procedures; and implement the new calibration data in each power range and OTAT protection channel.' In addition, SR 3.3.1.9 must be coordinated with other BOL power ascension testing and plant chemistry activities with consideration for fuel limitations, such as power ramp and rod withdrawal rates. An allowance of 7 days is also consistent with the time allowance for performance of Axial Flux Difference (AFD) surveillance in ITS SR 3.2.3.1, which requires calibrated excore channel AI indications based on the cycle-specific core power distributions. The 7 day allowance provides sufficient time for '
surveillance perfonnance, including excore channel Al normalization and the subsequent               j performance of the AFD surveillance, without placing unwarmnted duress on the plant operating,       I maintenance and engineering staffs. Finally, this allowance is consistent with the allowance         i I
provided in the approved Vogtle ITS.
Based on the above discussions, as supplemented by the revisions to FNP ITS Bases, DOC NOS.          j 84-A, 86a-L and 88a-M, and JFD NOS.10,11 and 12a, SNC has determined that the proposed surveillance frequency (18 months) and allowance (7 days) for the cycle-speciSc incore/excore cross-calibration are appropriate for FNP.
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Page 10 of 22                                            :
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p       '
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SNC Resperse to NRC RAI Related ta Ch pter 3.3 l.
i NRC Question No. 3.3.1-10 '

==Reference:==
: ITS 33.1, Action P.1 JFD #6 l           Justification for bypass allowance Note included with proposed action P.1 requires confumation that WCAP-10271 modeled this function for the bypass allowance during the testing of other channels, t
Comment: - Provide a statement confirming that WCAP-10271-P-A included the bypass allowance for test as proposed for ITS Action P.I.
SNC Response No. 3.3.1-10 Review of WCAP-10271-P-A and WCAP-10271 Supplement 1-P-A indicates that there is no approved provision in either of these Topical Reports for testing Reactor Trip System (RTS) .
Tuitine Throttle Valve closure channels with one or more inoperable channel (s) in " bypass."
Herefore, to maintain consistency with the Farley CTS and WCAP-10271, which is applicable to -
Farley, the proposed ITS Table 3.3.1-1 change that added Condition P has been revised by deleting the Note, which pertained to the bypass allowance. In addition, the revision clarifies that Condition P is applicable to "One, two or three Turbine Throttle Valve Closure chmanel(s) inoperable," versus "One or more Turbine Throttle Valve Closure channels inoperable."
Rese revisions are reflected in the STS and STS Bases markups. In addition, DOC NOs. 37-L, 63-A and 64-M, and JFD-6 have been revised.
NRC Question No.3.3.1-11 ~
The NRC Staff deleted this question via E-mail from Carl Schulten dated March 11,1999.
NRC Question No. 3.3.1 12

==Reference:==
SR 3.3.1.8 JFD 13 Proposed changes eliminate SR 3.3.1.8 COT testing for IRM channels upon reducing power below P-10. Proposed changes are inconsistent with TS operability requirements. The STS requires that these instruments be operabic for startup and during those times that the reactor is in the applicable modes i.e., upon reducing power below P-10 from above P-10. IfIRM channels have a current surveillance as permitted by SR 3.0.2 then the surveillance is not due upon reducing power below P-10. The STS Bases for SR 3.3.1.8 cite the importance of STS testing:
                    "Ihe Frequency of"4 hours after reducing power below P 10"(applicable to intermediate and power range low channels) and "4 hours after reducing power below P-6"(applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. He Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. He MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this Page 11 of 22

SNC R:sp=se to NRC RAI Relat:d to Chrpter 3.3 surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-10 or < P-6) for periods > 4 hours."
Comment: Provide additionaljustification for proposing a deviation from the STS. Justify why this proposed change is not generic. Explain in detail why the basis provided in the STS Bases is not applicable to Farley.
SNC Response No. 3.3.1-12 The proposed FNP ITS surveillance SR 3.3.1.8, as applied to NIS Intermediate Range (IR) protection system channels, excludes performance of function testing (COT) below P-10 based on the CTS, which have no such requirement. Based on further evaluation, SNC determined that FNP should conform to the STS, which allows 4 hours to perform the IR COT when reducing power below P-10 and every 92 days thereafter while in the mode of applicability. In conformance with the Staffs comment, ITS SR 3.3.1.8 and the Bases for SR 3.3.1.8 will be revised consistent with the STS, and JFD-13 will be revised. In addition, new DOC 86b-M has been added to discuss this change to the CTS.
NRC Question No. 3.3.1-13

==Reference:==
JFD 14 Proposed changes include combining the 18 month Channel Calibration SRs 3.3.1.10 and 3.3.1.11 into a single SR with 2 notes that are not applicable to all functions which require this test.
Comment: This is another unique and innovative proposed change to the STS. Remove these changes from the proposed ITS. Such changes require an approved TSTF before implementation in the Farley TS.
SNC Response No. 3.3.1-13 Based on further review, SNC confirmed that the proposed ITS changes are consistent with the Farley RTS design, calibration practices, setpoint uncertain +y calculations and CTS calibration requirements. The Farley ITS deletes STS RTS surveillance 3.3.1.12 and combines STS RTS surveillances 3.3.1.10 and 3.3.3.11 and the associated notes into one 18 month channel calibration surveillance. Thejustifications for these STS deviations are provided in JFD-14 and JFD-15.
The surveillance requires periodic channel calibration, and the notes clarify that neutron detectors are excluded from and that time constants are included in the calibration. While the surveillance is applicable to most RTS functions, the notes are not applicable to all of the specified RTS functions (e.g., pressurizer high pressure and steam generator low-low level); however, both notes can be applicable to a given function (e.g., power range high positive rate and OTAT). The time constant note must also be applied to dynamic time constants that are not explicitly listed in the Technical Specifications (e.g., pressurizer low pressure and RCP undervoltage). 'Ihe STS Bases clarify that the time constant note should be applied "where applicable." This interpretation is also appropriate for the neutron detector note. This approach is also consistent with the STS assignment ofjust one 18 month channel calibration survcillance with a single note for all ESFAS Page 12 of 22

rg                                                                                                            1 4
I SNC Resp =se to NRC RAI Related to Chapter 3.3 l-            ..
I         functions in STS ESFAS surveillance 3.3.2.9, whether the note was applicable to a given ESFAS function'or not. Given that these notes should only be applied if applicable and that ITS suiveillance 3.3.1.10 is consistent with the Farley design, calibration practices, setpoint calculations and CTS, the proposed change (i.e., STS deviation) is acceptable and the supportmg justifications do not require revision.
L'
       . NRC Question No. 3.3.1-14                                                                               !

==Reference:==
JFD 16 SR 3.3.1.13 3 i
       . The ITS proposes to change ITS SR 3.3.1.13 frequency to " Prior to exceeding P-9 after reactor startup" from " Prior to reactor startup."
Comment: De staff recommends modifying this proposed Frequency to read " Prior to exceeding P-9 dunng (emphasis added) reactor startup." The intent of the SR is to complete the test prior to entering the applicable conditiors.-
SNC Response No. 3.3.1 14 ITS Surveillance Requirement 3.3.1.13 converts the CTS turbine trip signal RTS channel A-*ia-I estst into a TADOT. De proposed change did not alter the test frequency of every reactor startup, nor did it alter the requirement to perform the test before exceeding P-9 during power ascension if not performed in the previous 31 days. Based on further review, in consideration of the NRC Staff review comment and comments made during the SNC-NRC meeting held on April 19-20,1999, SNC determined that the proposed change to CTS RTS Table 4.3 1 Note 9 should be revised. In addition, the ITS Bases will be revised.
De proposed change, consistent with TSTF-311, will require performance of the subject surveillance prior to exceeding the P-9 interlock whenever the unit has been in Mode 3, if not performed in the previous. 31 days. SNC concurs with the Staff determination that "[t]he intent of the SR is to complete the test prior to entering the applicable conditions."
The proposed change is supported by the Farley RTS design and the safety analyses, and it is             1' consistent with the CTS and STS operational mode of applicability (i.e., Mode 1 above P-9). De revised ITS SR 3.3.1.13 Frequency will read, " Prior to exceeding the P-9 interlock whenever the unit has been in Mode 3, if not performed in the previous 31 days " nis allows flexibility of test performance at any time prior to exceeding the P-9 interlock. The associated ITS Bases revision will reflect this change. In addition, the !TS Bases will be revised to clarify that the turbine must
       - be unlatched to verify the trip signal indication and that the turbine must be latched to reset the      ,
trip signal indication. The Bases clarifications are consistent with the Farley RTS design and the      l current surveillance procedures.                                                                        )
       ' As a rehult of these turbine trip RTS surveillance and Bases revisions, DOC NO. 93-LA has been revised and JFD-16 has been deleted.                                                                    l NRC Question No. 3.3.2-1                                                                                 j

==Reference:==
 ;ITS SR 3.3.2.9 STS SR 3.3.2.10                                                                         ,
CTS 4.3.2.3 DOC 4d-A
                                                    - Page 13 of 22 i
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a l'
SNC Resp:nse ts NRC RAI Rel:ted to Chrpt:r 3.3 A Note is added to the Response Time Testing requirement of CTS 4.3.2.3 which allows the testing of the turbine-driven auxiliary feedwater pump [to be] delayed for 72 hours after T,, is 547 *F ensuring sufficient steam pressure exists to support the test performance. This allowance (discussed in DOC 4d-A) in ITS SR 3.3.2.9 is based on the allowable out of-service time for the pump itself. He corresponding STS SR 3.3.2.10 allows only 24 hours for this delay after Steam
     . Generator pressure is "1000 psig." It is unclear how the out-of-service time for the auxiliary feed pump is a basis for determining this delay time. It seems the delay time should be based on the system capability or restrictions on reaching a stable steam pressure, rather than the allowed outage time of the pump, and the stability of steam pressure should be based on actual steam pressure instead of gT .
Comment:          Provide additional discussion andjustification for the deviation from the STS based on specific system design.
      ~ SNC Response No. 3.3.2-1 After further evaluation, SNC has determined that the ESFAS response time testing notation associated with ITS surveillance 3.3.2.9 should be revised to conform with the format of the STS.
De revised note will read, "Not required to be performed for the turbine-driven AFW pump until 24 hours aAcr SG pressure is 21005 psig." The time duration of 24 hours for performance of turbine driven auxiliary feedwater (TDAFW) pump response time testing is acceptable based on plant operating and testing experience. De steam generator pressure requirement of 21005 psig corresponds to the Reactor Coolant System no-load, hot standby, Mode 2 operating T, of 547
       *F. His Mode 2 requirement is consistent with the Farley safety analyses, which credit
     . automatic start of the TDAFW pump by RCP undervoltage as a primary engineered safety feature in the small break LOCA (SBLOCA) analyses. No other Farley safety analyses credit the TDAFW pump as a primary engineered safety feature. In that the SBLOCA analyses only require automatic ESFAS response in Modes 1 and 2, automatic TDAFW pump start by RCP undervoltage is only required in Modes 1 and 2. De no-load RCS temperature is also consistent with the safety analyses initial condition assumption for 0 % RTP (i.e., hot zero power). Since the no-load RCS Tgcorrelates directly to the no-load steam generator pressure,1005 psig is an appropriate reference condition for TDAFW response time testing. However, the ITS SR 3.3.2.9 notation does not preclude TDAFW response time testing at lower steam generator pressures.
De ESFAS STS and STS Bases mark-ups have been revised to reflect the above change and discussions. In addition, supporting justifications in DOC NO. 4d-A and JFD-6 have also been revised.
During this evaluation, SNC also noted that the statement of specified required channels was inconsistent with the FNP design and that response time testing and quarterly setpoint verification surveillance requirements had been omitted for ESF Function No. 6.d, "Undervoltage Reactor Coolant Pump," on Table 3.3.2-1. De findings associated with required channels and quarterly setpoint verification also apply to RTS Function No.12, "Undervoltage RCPs," and Function No.
13, "Underfrequency RCPs," on ITS Table 3.3.1-1. Therefore, included in the above revisions are corrections to RCP Bus UV & UF required channels and surveillance requirements. Dese revisions (i.e., corrections) are shown in the RTS/ESFAS STS and STS Bases mark-ups. ESFAS Instrumentation CTS justification DOC NO. 32-A has been revised, and STS deviation justifications JFD-6a and JFD-6b have been added. RTS Instrumentation CTS justification DOC NOS. 39-A and 91-A have been revised, and STS deviationjustification JFD-13b has been added.
Page 14 of 22

                              ' SNC Resp:nse ta NRC RAI Rellt:d ta Ch:pt r 3.3 i
NRC Question No. 3.3.2 2 -

==Reference:==
 . ITS SR 3.3.2.8 STS SR 3.3.2.6 f                      DOC 79M De CTS does not specify Slave Relay Testing performed on the ESFAS Automatic Actuation
    . Logic and Actuation Relays. ITS SR 3.3.2.8 adds this surveillance requirement consistent with the STS, except on an 18 month interval. STS SR 3.3.2.6 requires these tests on a 92 day interval. His Surveillance Test Interval deviation from the STS (discussed in DOC 79M) is justified as being per the plant FSAR. It is not clear why this test is not required in the CTS if the requirement is contained in the plant FSAR.
Comment: Provide additional discussion andjustification for the 18 month surveillance interval, based on current licensing basis and specific plant design. Also, provide discussion for the justification for deleting STS SR 3.3.2.6 and replacing it with ITS SR 3.3.2.8, requiring             {
renumbering of the STS SRs in the ITS.                                                                   J l
SNC Response No. 3.3.2-2                .
De FNP R'Ili and ESFAS design employs the logic, reactor trip, and ESF actuation circuits furnished with the Westinghouse Solid State Protection System as specified by the functional system requirements and logic diagrams. De slave relays provide the electrical interface between the automatic and manual ESF actuation logic and the ESF equipment (i.e., pumps, .
    ; valves, etc.). Here is no explicit requirement for a periodic SLAVE RELAY TEST in the Farley CTS. Nevertheless, in that the slave relays are electrical components in the ESFAS signal path, the relays should be periodically tested to demonstrate operability. In the Westinghouse SSPS design, the slave relays are tested in part during the MASTER RELAY TEST and in-part or fully tested during the SLAVE RELAY TEST. The SLAVE RELAY TEST assumes that the relays are tested using the Safeguards Test Cabinet circuits. However, slave relays can also be tested during        i other activities such as response time testing and integrated safeguards testing. He current Farley practice is to test the slave relays in both tnuns every 18 months. Normally, slave relay testing is performed during refueling outages to minimize the potential for plant transients and unnecessary challenges to plant equipment. De 18-month testing requirement, including a description of the SLAVE RELAY TEST, is included in Chapter 7.3 of the Farley FSAR, and therefore, it is a past of the current design / licensing basis, ne technical basis for the Farley.
specific test frequency, which includes relay reliability, failure analysis and operating experience, is provided in WCAP-13877, " Reliability Assessment of Westinghouse Type AR Relays Used as               i SSPS Slave Relays WOG Program MUHP-7040." WCAP 13877 is listed in FSAR Chapter 7.3                       l as Reference No.12.
ESFAS DOC NOJ 79-M and JFD-5 provide the justifications for the subject ITS change. %cse Edocuments do not warrant revision. However, the STS Bases markup for proposed ITS SR3.3.2.8 has been revised to clarify that slave relay testing is normally conducted during refueling to           ;
minimize the potential for plant transients and unnecessary challenges to plant equipment.               ,
l With regard to the ESFAS surveillance numbering, STS SR 3.3.2.6 and ITS SR 3.3.2.8 are both identified as the SLAVE RELAY TEST. STS SR 3.3.2.6 is a "backeted" 92-day test. ITS SR                   i 3.3.2.8 is an 18-month test. Consistent with the STS format, the Farley ESFAS SLAVE RELAY                ,
TEST was placed in the appropriate position for an ascending order sequence based on test frequency and renumbered.
4 Page 15 of 22 i
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SNC Respoise to NRC RAI Rited 13 Chapt:r 3.3 i
NRC Question No. 3.3.2-3

==Reference:==
ITS Table 3.3.2-1 footnote (g)                                                              .
1 STS Table 3.3.2-1 footnote (j)
CTS Table 3.3-1                                        ,                              f DOC 28L                                                                                !
The CTS Table 3.3-1 Applicability for the Steam Generator Water Level - High-High Function of Turbine Trip and FW Isolation is revised in the ITS Table 3.3.2 1 footnote (g) to exempt Applicability of the Function when all MF lines are isolated by either a MF Stop Valve, an            1 MFRV and associated bypass valve, or by a closed manual isolation (Ref DOC 28L). STS Table 3.3.2 1 footnote (j) provides this exemption only when "all" stop valves, MFRVs and             )
associated bypass valves are " closed and de-activated," or " isolated by a closed manual valve "
Justification for providing the exemption by "either" a stop valve or an MFRV is based on the         j additional redundancy ofisolation capability in the FNP design. But the requirement to ensure the stop valve or MFRV and associated bypass valve is "de-activated" when isolation is not ensured by a closed manual valve is omitted in the corresponding ITS Table 3.3.2-1 footnote (g).
Provide additional discussion andjustification for the STS deviation based on specific plant design.
Comment: Provide additional discussion andjustification for the STS deviation based on specific plant design.
  - SNC Response No. 3.3.2-3                                                                           ,
Based on further reviews, SNC has determined that ITS Table 3.3.2 1 footnote (j) does not provide any real benefit to sustaining Farley unit operation for any conceivable P-14 logic or channel failure in Mode 2. For example, a postulated logic failure of the Turbine Trip and             j Feedwater Isolation function may require entry into a more restrictive LCO because some logic and actuation cir.:uits are shared with the ESFAS Safety Injection function, which requires the automatic actuaton logic to be operable in Modes 1,2, and 3.
Therefore, footnote (g) will be deleted from the ITS Table 3.3.2-1. This action is consistent with the CTS Table 3.3-1 which does not currently provide for an exception for the P-14 function in         i Mode 2. Since the footnote (g) will be deleted from ITS Table 3.3.2-1, the NRC comment no               i I
longer applies.
The STS and STS Bases mark-ups and associatedjustifications have been revised to reflect these changes. In addition, the STS Bases mark-ups have been revised to emphasize the safety function afforded by safety injection actuation of feedwater isolation and SGFP trip.
NRC Question No. 3.3.2-4

==Reference:==
JFD 1 ITS Condition J l
Proposed changes include replacing the STS 48 hour completion time with CTS " prior to next required TADOT."
Comment: - The proposed completion time change is for the AFW pump start on trip of all MFW pumps. The STS 48 hour completion time is based on application of WCAP-10271 l
Page 16 of 22

O                             . SNC Response ts NRC RAI Rel:t:d 13 Chapter 3.3 analyses. Provide additional discussion to justify deviating from WCAP-10271 when this WCAP is otherwise adopted in ITS.
SNC Response No. 3.3.2-4 '
      . WCAP-10271-P-A, Supplement 2, Revision 1, " Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," did not evaluate the required action completion time for the Motor-Driven AFW (MDAFW) Pump automatic start by trip of both Main Feedwater Pumps. De ESFAS Condition J required action complete time of 48 hours in STS was added as ESFAS Table 3.3-3 Action No.19 in NUREG-0452, Revision 4, .
        " Standard T-h=6-1 Specifications for Westinghouse Pressurized Water Reactors," Fall 1981.
SNC can not determine the basis for the 48 hour completion time. However, NUREG 0452, Revision 4, is not part of the FNP licensing basis; therefore, it should not be applied to Farley.
As specified in the Westinghouse functional system requirements and functional logic diagrams, the Farley design includes an automatic start of the MDAFW Pumps upon trip of both Steam Generator Feed Pump (SGFP) turbines. His design feature is an anticipatory AFW start signal.
De signal is not credited in the Farley safety analyses as a primary ESFAS signal, nor is it credited for diversity.L (With regard to postulated loss of feedwater events, manual operator action is credited for diversity; see Section 5.2 of WCAP-7306, " Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors," April 1969.) As such, these circuits are not required to be safety-grade or single-failure proof. De as-built configuration meets the design requirements by using 4 channels, i.e.~,2 channels per pump or 1 channel per limit switch. Each SGFP turbine HP and LP steam supply stop valve uses I limit switch to sense closed valve position. Each limit switch channel provides inputs to the 4-out of-4 logic circuit for each MDAFW pump. SGFP trip signals result in low EHC fluid pressure and closure of the HP and
      . LP steam supply valves. Should both feed pumps trip, the AFW motor driven pumps are automatically started The automatic MDAFW Pump start by "frip of Main Feedwater Pumps"is listed on ESFAS Table 3.3-3 as Functional Unit No. 6.c in the Farley CTS. He function is required to be operable in Mode 1, and it is functionally tested prior to reactor startup. There is no requirement to shutdown the unit or reduce power if the function is inoperable. Dese Farley CTS requirements reflect the back-up/ anticipatory nature of this function and the fact that no safety analyses credit automatic AFW startup by tripping of the SGFP's. Therefore, to be consistent with the current Farley licensing basis, the ITS will retain the requirement to ensure that this signal is operable prior to startup and not add any new action / completion time requirement that results in unit power reduction or Mode changes.
      > Justifications provided by DOC NOS. 34-M,54-L and 55-A explain the basis for converting the Farley CTS to ITS, and JFD-1, JFD-21 and JFD-22 provide the basis for the STS deviations.               ;
        %ese references do not require revision. However, the ITS Bases will be enh=am4 by clarifying
,        that "this ESF function is not credited for diversity, and its electrical circuits are not required to be safety-grade." (See revised STS Bases mark-up.) ne proposed ITS Bases already state that
       ' the " function is not relied on in any safety analyses as the primary actuation signal to initiate the AFW pumps. . . ."
NRC Question No. 3.3.2-5

==Reference:==
JFD 3                                                                                        4 ITS Condition K                                                                         l l
l.
Page 17 of 22 '
l-                                                                                                               !
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I SNC Resporse to NRC RAI Rel:td ts Chapt:r 3.3                               I Proposed changes add "or more" language to the condition for inoperable P-11 and P-12 interlock
                                             '      ~
channels.
Comment: This proposed allowance avoids entry into LCO 3.0.3 for channels that become
      . inoperable subsequent to the first inoperable channel of the function. Provide discussion justifying a 1-hour completion time for a second inoperable channel in a two-of-three actuation     ;
logic.                                                                                             j SNC Response No. 3.3.2-5 Based on fusiher review, SNC determined that certain CTS limiting conditions for operation and     l surveillance requirements for the ESFAS permissives P-4, P-11, and P-12 had not been properly implemented during the STS conversion to ITS. Therefore, SNC has revised the ITS Conditions and Surveillance Requirements for ITS Function No. 8,"ESFAS Interlocks," on Table 3.3.2-1 to be consistent with the current Farley licensing basis as defined by CTS, current plant practices,
      - and applicable setpoint vaccatainty calculations. 'Ihese revisions require changes to the STS and Bases mark-ups, DOC NOS. 4b-A and 50-M, JFD-3, JFD-7, and JFD-22. Also, new DOC NOS.
98-A and 99-M and JFD 22a were added. In that these revisions modified ITS Condition K (based on Farley CTS Table 3.3-3 Action Nos.13,19 and 20, the ESFAS design, and STS Condition L), the above comment is no longer applicable.
NRC Question No. 3.3.2-6

==Reference:==
JFD 6 ITS SR 3.3.2.9 Proposed change replaces the 24-hour STS time-period for performing the AFW pump response time test after reaching appropriate test conditions with 72 hours.
Comment: Thejustifications presented in the JFD rely on comparisons of the 72-hour test period with the allowance in the ITS for an inoperable AFW pump. Providejustification that cites reasons explaining why the 24 hours is an insufficient time period to accomplish required testing.
SNC Response No. 3.3.2-6
        'Ihe SNC response to this NRC Staff comment is included in the SNC response to Question No.
3.3.2-1, wherein, SNC stated that the Farley ITS notation for SR 3.3.2.9 would be revised to be consistent with the format of the STS.
NRC Question No. 3.3.4-1                                                                            ,
l'

==Reference:==
ITS 3.3.4, Remote Shutdown System, Condition C and Condition B Note CTS 3/4.3.3.5, Remote Shutdown Instrumentation                                    j~
                       . DOC 10a-M, & JFD 4 -
Neither CTS 3/4.3.3.5 , Remote Shutdown Instrumentation, nor STS 3.3.4, Remote Shutdown
      ~ System, contain explicit Conditions or Required Actions related to the Source Range Neutron Flux Function. CTS 3/4.3.3.5 does not contain the Source Range Monitor as part of the Remote        .
Shutdown System, and STS 3.3.4 considers the Source Range Neutron Flux Function the same as         l other Remote Shutdown Functions and therefore it falls under the same Conditions and Required
      . Actions as the other Functions. 'Ihe ITS takes exception to the STS requirement to shutdown the l
Page 18 'of 22 i

5 SNC Response is NRC RAI Rel:ted 13 Ch ptir 3.3 L       unit if the Source Range Neutron Flux Function is inoperable for more than 30 days. Instead the . I ITS proposes a report be submitted to the NRC in 30 days describing an alternate means of ensuring the reactor remains shutdown in the event of a control [ room] evacuation.
Comment:          Since this change is neither in conformance with the STS nor the CTS, it is a      ,
Beyond Scope change which requires NRC Tech Staff review. Since this function is required to        I be operable to maintain the plant in shutdown, as part of the Remote Shutdown System, it is not      l L       clear why the STS Condition and Required Actions are not adopted in the ITS. 'Ihejustification given, based upon a plant unique design is not clear, Explain the basis for the ITS proposal and
       ' include a discussion of the alternate means available to ensure that the reactor remains shutdown in the event of a control room evacuation.
SNC Response No. 3.3.41
        'Ihe instrumentation listed in ITS Table 3.3.4-1, including the source range neutron flux instrumentation, is consistent with the FNP current licensing basis as discussed in FSAR section 7.4 This section identifies instrumentation necessary for ccmpliance with 10 CFR 50, Appendix A, GDC 19. SNC has agreed to adopt TS requirements for source range neutron flux in the FNP l       ITS. However, the FNP current licensing basis for source range neutron flux consists ofone channel of source range indication independent from the source range instrumentation used for          i I
reactor protection. SNC is unwilling to adopt requirements that could result in shutdown of a unit due to a single inoperable channel of source range indication on the hot shutdown panels. SNC believes that such a requirement is unnecessary for the following reasons.
The likelihood of a control room fire that results in evacuation concurrent with an inoperable channel of source range indication is small. A significant fire in the main control room is not l       considered a credible ever+. since it is continuously occupied, power circuits are minimized and       i l        combustibles are limited. Even if such a fire were to occur, lack of source range indication would    j not prevent achieving and maintaining a safe hot shutdown condition. Abnormal Operating                i Procedures for responding to fires in the main control room direct operators, in part, to trip the l        reactor and realign charging pump suction to the refueling water storage tank prior to evacuating the control room. 'Ihis ensures that control and shutdown rods are inserted and that a source of borated water is avadable. Once outside the control room, operators are directed to ensure that the reactor remains subcritical by sampling for RCS boron concentration periodically and performing shutdown margin (SDM) calculations. In addition, adequate RCS SDM must be achieved prior to plant cooldown. The instrumentation provided on the hot shutdown panels is           i sufficient to allow for shutdown margin calculations without having to rely on source range           ;
indication. 'Iherefore, the proposed requirements for source range indication are sufficient and      i consistent with FNP current licensing basis.
l NRC Question No. 3.3.4-2 1

==Reference:==
ITS 3.3.4, Remote Shutdown System l
ITS Table 3.3.4-1, Remote Shutdown lastrumentation and Controls                      ;
!                         CTS Table 3.3 9, Remote Shutdown Monitoring Instrumentation                          l I
DOC 11-M, & JFD 3 ITS Table 3.3.4-1 contains the Remote Shutdown Monitoring Instrumentation contained in CTS Table 3.3-9 and some of the Remote Shutdown Instrumentation and Controls listed in STS Table 3.3.4-1, but not all of the STS lis:ed Insuumentation and Controls.
i:
Page 19 of 22-l L

p SNC Respo se to NRC RAI Relat:d 13 Chapter 3.3
[
Comment: ITS Table 3.3.4-1 does not include the Manual Reac;or Trip, Reactor Trip Breaker Position, and Source Range Neutron Flux functions listed in STS Table 3.3.4-1; why not? The ITS Table 3.3.4-1 does not use a format similar to either the STS or CTS tables; why not? %e ITS Transfer and Control Circuits do not list the required number of channels; why not? ITS Transfer and Control Circuit Function 5 should be defined either on the table or in the Bases; it relies too much on opemtor interpretation.
L SNC Response No. 3.3.4-2 STS Table 3.3.4-1 is furnished with a Reviewers Note that states that the STS table is for illustration purposes only. De FNP ITS contains those functions that are consistent with the -
FNP cummt licensms basis, including requirements for source range indication. De FNP hot
,            shutdown panels are not equipped with manual reactor trip switches or reactor trip breaker .
           - position. De Source Range Neutron Flux function (Gammametrics) is included as Function 7
;'          under Monitoring Instrumentation.
I l           The ITS format for Table 3.3.4-1 is more operator friendly than the STS format in that the ITS l            format separates indication from transfer and control circuits whereas the STS format mixes
;-           indication with control. However, with respect to the required number of channels for transfer
!           and control circuits, SNC will revise the FNP ITS submittal to specify one required channel for each transfer and control circuit function.
l           With respect to ITS Transfer and Control Circuit Function 5, the Bases already adequately l~          address the systems necessary for meeting this requirement. The LCO section of B 3.3.4 discusses safety support systems including service water, component cooling water, and offsite power, including the diesel generators. Taking each function under Transfer and Control l          . Circuits, offsite power (including diesel generators) is required to support each of the specified functions, and service water and/or component cooling water is required to support the charging system and auxiliary feedwater system.                                                             J NRC Question No. 3.3.5-1

==Reference:==
ITS 3.3.5, LOP DG Staat Instrumentation, Conditions A, B, & D and Required Action B.1 STS 3.3.5, LOP DG Start Instrumentation p                           ' JFD 2 STS 3.3.5 is significantly revised in creating ITS 3.3.5, in order to accommodate an existing commitment (as described in JFD 1). His is acceptable. In revising STS 3.3.5, the wording for Conditions A & B, and Required Action B.1 has been changed.
{
i
!.           Comment: The proposed changes to Conditions A & B should be written to take advantage of            !
             " Separate Condition entry" which allows these conditions to be entered for more than one           i function at a time. He channels provided in Table 3.3.5-1 support " bus" operability, not " train"  !
operability.' Table 3.3.5-1 should be revised to indicate " Required Channels" in the column
           . header and the Functions in the table should be identified as "3 per bus" for Functions 1 and 2 and."1 per bus" for Function 3. Given the use of a table for identifying required LOP DG Start functions, the STS format for applicable conditions is best applied with the following changes:     2
                          ,                            Page 20 of 22-1:.                                                                                                            ;

SNC Response is NGC RAI Rel:ted to Chapttr 3.3 Add a " Note" to Condition A and to Condition B: "Only Applicable to Functions 1 and 2" Condition A, "One channel inoperable." '
Condition B,      "Two or more channels inoperable."
Delete th: use of a " train" reference in Required Action B.1" Add a " Note" to Condition D, "Only Applicable to Function 3" SNC Response No. 3.3.5-1 ne ITS submittal will be revised to include a " Note" for Condition A and B: "Only applicable to Functions 1 and 2." Condition A will be revised to read "One or more Functions with one channel per train inoperable." Condition B will be revised to read "One or more Functions with two or more channels per train inoperable." He reference to trams is appropriate because there is a 4.16 kV bus per tmin and each bus / train is equipped with 3 channels ofinstrumentation.
    . Condition D will be revised to add a " Note": "Only applicable to Function 3."
With the above revisions, it will be clear as to the applicability of each condition, and there should be no confusion as to the applicability of separate condition entry.
NRC Question No. 3.3.5 2

==Reference:==
ITS 3.3.5, LOP DG Start Instrumentation, SR 3.3.5.1 Note 2                                 ,
STS 3.3.5, LOP DG Start Instrumentation                                               I CTS Table 4.3-2 Notes 3 & 4 JFD 6, and DOCS 90-A & 91-A In the performance of the TADOT required by the CTS, fmal trip actuation is excluded.
i Comment: The ITS excludes final trip actuation by note 1. He ITS includes an additional note 2, which is not in either the CTS or the STS, that states that setpoint verification is not required. Why is note 2 included?
SNC Response No. 3.3.5-2 Note 2 (i.e., setpoint verification is not required) is included as part of the FNP current hcensing   !
basis (see Note (4) to Table 4.3-2 (CTS page 3/4 3-37)). As stated in ESFAS Section 3.3.2, DOC 90-A, CTS Note (4) excludes setpoint verification for the monthly channel fut.ctional test by          j describing the test as removal ofinput voltage and verification of relay operation. Herefore, Note 2 is added in the ITS to make it explicitly clear that setpoint verification is not part of the ITS TADOT.
NRC Question No. 3.3.8-1

==Reference:==
ITS 3.3.8, Condition C JFD 6
      %e ITS proposes a the Note to Condition C, "Only applicable to Functions required OPERABLE by Table 3.3.8-1 during movement ofirradiated fuel assemblies in the spent fuel pool room."
l Comment: ' Remove this proposed aote from the ITS. Condition C as written includes the same provisions as proposed by the note.
Page 21 of 22

SNC R:sprnse to NRC RAI Rtl-ted 13 Chrpt:r 3.3 SNC Response No. 3.3.8-1 The proposed Note to Condition C serves a useful purpose. Consider the following scenario. A unit is in Mode 1, and movement ofirradiated fuel assemblies is underway in the spent fuel pool room.' Now suppose that one train of automatic actuation logic and actuation relays is determined to be inoperable. Per Table 3.3.8-1, the automatic actuation logic and actuation relays are required to be operable in Modes 12,3 and 4, but not during movement ofirradiated fuel assemblies in the spent fuel pool room. Therefore, in the absence of the proposed Note to Condition C, Condition C could become applicable and force an unnecessary suspension of fuel movement.
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i Page 22 of 22                                            l

i L'
ATTACHMENTII SNC Response to NRC Request for Additional Information Related to Conversion to the Improved Technical Specifications - Chapter 3.3 Associated Package Changes Grouped by RAI Number

Associated Package Changes for RAI- 3.3.1-1

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.3 - Instrumentation CTS 2.2 LIMITING SAFETY SYSTEM SETTINGS CTS 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION FNP ITS 3.3.1 REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION DOC NQ                       SHE                                     DISCUSSION the bases for the low flow trip function (described in previous DOC).
6                    A     The P-7 RTS Interlock function setpoints are reorganized consistent with the presentation of this information in the STS. The P-7 interlock does not have setpoints ofits own but relies on the input from RTS Interlocks P-10 and P-13. The STS organizes all the setpoints affecting the P-7 interlock under the appropriate interlock (P-10 or P-13) from which the input to P-7 originates. P-7 is revised to conectly indicate that no setpoint is associated with it. Since the setpoints of P-10 and P-13 remain unchanged, the actuation of P-7 also remains unchanged. This re-organization of the setpoint table is made only to conform with the presentation and format of this information in the STS and does not introduce a technical change to the CTS. Therefore, this change is considered admmistrative.
7                    A     Notes 4 and 5 associated with the RTS interlock P-13 on CTS Table 2.2-1 are deleted. These notes contain guidance for determining the P-13 setpoint equivalent impulse pressure during initial plant startup testing. The notes are no longer applicable as initial startup testing is completed. Since the notes no longer apply, deletion of these notes may be considered an administrative change.
8                     A     The CTS Table 2.2-1 is revised by the addition of the Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms Function consistent with the STS. The STS provides a separate line item for these reactor trip breaker components in the corresponding STS Table 3.3.1-1. The STS further separates the applicable Actions and surveillances for these components as well. The Action and surveillance requirements for this separate line item are derived from the existing reactor trip breaker requirements. As such, this change represents a re-organization of the setpoint table to conform 3
y[                       with the STS format and presentation of the reactor trip breaker requirements. Therefore, the introduction ofthis new line item on the table is considered an administrative change.
9                     L     The CTS Table 2.2-1 Notes 1 and 2 containing the overtemperature and overpower delta T equations are revised to be consistent with the format and presentation of the corresponding STS equations, where appropriate,
        .                 Chapter 3.3                                                  E2-4-A                                      May,1999 1

FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.3 - Instrumentation CTS 2.2 LIMITING SAFETY SYSTEM SETTINGS CTS 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION FNP ITS 3.3.1 REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION DOC NQ        SHE                                       DISCUSSION and the applicable FNP specific safety analyses. The revision consists of changing ti, ti , and T3 from an equality to an inequality. The direction of g
conservatism identified by the proposed inequalities is consistent with the p.V             modeling of these parameters in the applicable FNP safety analyses. Setting time constants in the direction of conservatism will ensure conservative protection function response with respect to the modeling of these functions in the design basis accident analyses.
10        A      The overtemperature and overpower delta T equations in Notes 1 and 2 on CTS Table 2.2-1 are revised, where appropriate, by the deletion of descriptive information consistent with the format and presentation of this information in the STS (Notes 1 and 2 of Table 3.3.1-1). Specifically, the l
CTS notes 1 and 2 for overtemperature and overpower delta T setpoints I                     contain descriptions of the dynamic compensation terms used in the setpoint equations. These descriptions constitute detail beyond that included in the technical specifications for any other safety analysis calculations. In addition, this detail merely restates the information contained in the setpoint equations. Therefore, the CTS Table 2.2-1 notes are revised to be consistent with the STS by deleting the dynamic compensation descriptive detail redundant to the information contained in the setpoint equation. Since the information deleted does not reduce or change the CTS requirements for the ove temperature and overpower delta T setpoints (the equations and variables remain unchanged), this change is considered administrative and is made solely to conform with the presentation and format of this information in the STS.
I1        A    - The CTS Note 1 on Table 2.2-1 is revised to delete references to two loop operaten. The CTS contains many references to two loop operation; l_
however, FNP is not licensed for two loop operation and these references do not contain requirements that are applicable to FNP. Therefore, all references to two loop operation are being deleted. If two loop operation specific requirements are deleted, the title of 3 loop operation referring to the remaining requirements is also deleted, since only 3 loop operation is licensed at FNP. As these changes do not impact the cuttent applicable license requirements for FNP, they are considered administrative.
Chapter 3.3                                      E2-5-A                                                 May,1999

m FNP TS Conversion Enclosure 2 - Discussion of Changes to CTS Chapter 3.3 - Instrumentation CTS 2.2 LIMITING SAFETY SYSTEM SETTINGS CTS 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION FNP ITS 3.3.1 REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION DOC NQ       SHE'                                    DISCUSSION 12'      A     The definition for fi (AI) in Note 1 on CTS Table 2.2-1 is revised consistent with the similar expression in Note 1 of STS Table 3.3.1-1. The three components of this definition are revised to be expressed in equation form consistent with the fonnat and presentation of this information in the STS.
This change is not intended to introduce a technical change in the CTS requirements. The CTS requirements are simply re-stated in a different form to be consistent with the STS. As such this change is considered administnaive.
13       A     Notes 3 and 6 of CTS Table 2.2-1 refer to the allowable values for the overtemperature and overpower delta T functions. These Notes have been effectively incorporated into Notes 1 and 2 respectively, consistent with the format and presentation of this information in the STS Table 3.3.1-1. The l
3.'y\         Farley-specific values contained within the Notes are not changed. The Notes are re-organized solely to conform with Notes 1 and 2 of STS Table 3.3.1-1. Therefore, this change is considered administrative.
14       A     The CTS 3/4.3.1 LCO statement, Applicability, and Actions are revised consistent with the STS format and presentation of this information. The LCO statement is revised to eliminate the phrase "as a minimum". The LCO requirements for a system or component are the minimum requirements by definition of the term "LCO". Therefore, the CTS term "as a minimum"is not necessary and has been deleted. The LCO statement is also revised to address the STS Table 3.3.1-1 and the functions listed on that Table. The Applicability of CTS 3/4.3.1 is revised to refer to the STS Table 3.3.1-1 instead of the CTS Table. In addition, the CTS 3/4.3.1             !
I Actions are revised consistent with the STS. The Actions for CTS 3/4.3.1 are revised to be in the Condition, Required Action, and Completion Time format of all other STS Technical Specifications and the simple reference to CTS Table 3.3-1 in the CTS Action is replaced with the STS Condition A which states the condition of one or more Functions (on Table 3.3.1-1) with one or more inoperable channels or trams and also references the applicable Condition on STS Table 3.3.1-1. The STS Condition A, although more               ,
technical complete in stating the applicable condition, effectively              !
accomplishes the same thing as the simple CTS reference to the Table.            !
These changes are made to conform with the format and presentation of the Chapter 3.3                                   E2-6-A                                    May,1999   i l

FNP TS Conversion Enclosure 3 - Significant Hazards Evaluations Chapter 3.3 - Instrumentation III. SPECIFIC SIGNIFICANT HAZARDS EVALUATIONS CTS 2.2 LIMITING SAFETY SYSTEM SETTINGS CTS 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION
       /           FNP ITS 3.3.1 REACTOR TRIP SYSTEM (RTS) INSTRUMENTATION 3.T
  +                                                       2:L
: 1. Does the change involve a significant increase in the probability or consequences of an            ,
accident previously evaluated?      <
l       The proposed change involves changing the dynamic time constant equality expressions in           j the overtemperature and overpower delta T equations to more closely agree with the STS requirements and does not result in any hardware changes. The existing ti, t ,i and r3 equalities are replaced with inequalities to ensure the associated time constants are set in a conservative direction relative to the modeling assumptions of the safety analyses. The overtemperature and overpower delta T functions are not assumed to be an initiator of any I       analyzed event. The role of these instmment functions is to mitipte design basis accidents by tripping the reactor. The proposed change still ensures the system remains capable of providing the required mitiption function as described in the FSAR and that the results of '
the analyses in the FSAR remain bounding. Additionally, the proposed change does not              ,
impose any new safety analyses limits or alter the protection systems' ability to detect and mitipte design basis events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change involves changing the TS requirements for selected overtemperature and overpressure delta T dynamic time constants to more closely agree with the safety analyses and does not necessitate a physical alteration of the plant or changes in parameters goveming normal plant operation. The change is also consistent with existing overtemperature and overpower delta T scaling and calibration practices. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does this change involve a significant reduction in a margin of safety?
The proposed change, which revises the ti, T2 , and t3 dynamic time constant equality for the overtemperature and overpower delta T reactor trip functions to be more consistent with the STS requirements, does not involve a significant reduction in a margin of safety. The .           ;
proposed change will require the applicable time constants to be conservatively set with         j respect to the assumptions of the safety analyses. The ove temperature and overpower delta        !
T functions will continue to provide the required mitiption actuation as described in the         i FSAR and thereby assure the safety analyses limits are not exceeded. Therefore, the
. Chapter 3.3                                   E3-1-A                                     May,1999 j
i j}}