ML20203A755

Forwards Response to NRC 981214 e-mail Re Tech Specs Chapter 3.4,converting to Its.Proposed Chapter 3.4,addressing Editorial Changes,Omissions & Consistencies in Package Identified After Original Submittal Also Encl

ML20203A755
Person / Time
Site:	Farley
Issue date:	02/03/1999
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20203A758	List: ML20203A755 ML20203A771
References
NEL-99-0038, NEL-99-38, NUDOCS 9902100092
Download: ML20203A755 (20)

Text

- -

'Morey SIuthern Nuclear

,o t' resident Op rating Compa y tarley Project P.C. Box 1295

( Birmingham Alabama 35201 Tel 205.992.5131

[

SOUTHERN February.3, 1999 EnergytoServeYourWorM*

Docket Nos.: 50-348 NEL-99-0038 50-364 U. S.Nachar Regulatory Commission ATIN.: Document Control Desk Washmaton, D. C. 20555 0001 Joseph M. Farley Nuclear Plant Response to Request for Additional Information Related to Converei= to the i=aroved Tdair=1 Raari&=*ia== - Ch=ater 3.4 Ladies and Gentlemen:

By laters dated March 12,1998, and April 24,1998, and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) submitted the Farley Nuclear Plant (FNP) - specific Improved Tachar=1 Specifications (ITS) conversion documentation packages. The April 24, 1998 letter, which submitted the Clean-Typed copies of the FNP ITS, included an attachment

- which prwided hard copies of changes to the original submittal to correct minor editorial errors

' and inconsimeacies within the package. By later dated August 20,1998, SNC submitted an electronic copy of the Discussion of Changes (DOCS) and Significant Hazards Evaluations (SHEs) ==acintad with the ITS conversion. Included with that letter were hard copies of changes to the original subauttal to correct minor editonal errors and inconsistencies within the package.

By 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 copics of changes to the original subauttal to reflect the SNC responses to the RAI. NRC E-mail dated December 14,1998, requested SNC prwide additional information for Chapter 3.4.

Ah I providas the SNC responses to the NRC questions. /*d-=* II includes pmposed revisions to the previously subautted license amendment request related to this RAI, grouped by RAI number. Attachnent III providcs changes made to the Chapter 3.4 submittal to address editanal changes, omissions, and incr=aimaacies in the package identified aAct the original submittal.

Clean-typed copics 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.

080043 '~ TjD\

i 9902100092 990203  ?.

PDR ADOCK 05000348 i P PDR y p, ,

r

1 o I Page 2 U. S. Nuclear Regulatory Commission I

Mr. D. N. Morey states that he is a Vice President of Southern Nuclear Operating Company and is authonzod to execute this oath on behalf of Southern Nuclear Operating Company and that, to the best of his knowledge and belief, the facts set for*.h in this letter and attachments are true.

If you have any questions, please advise.

Pa%11y submitted, SOUIRERN NUCLEAR OPERATING COMPANY ft 71hY Dave Morey Sworn to andsubscribed before me t ish. day of SW)999 h)'  %$ () Notary Public My Commission Expires:

u

l. NO l _

WAS:r.d itsrai_2. doc Attachnu,r.ts cc: See next page.

t l Page 3 '

l U. S. Nuclear Regulatory Commission

! cc: U. S. Nuclear Re=da*w Commissiort W=hiantan D. C.

Mr. J. I. Zimmerman, Licensing Project Manager - Farley U. S. Nuclear Renulatory Ca==lesia Reaion II I Mr. L A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Faricy  ;

l Alabama Denartment of Public Health l Dr. D. E. Williamson, State Health Officer i

i i

i I

l l

j

\

I g.

l . i

? l 1

l l

l ATTACHMENTI l

SNC Response to NRC Request for Additional Information Related to Conversion to the Improved Technical Specifications - Chapter 3.4 I

i l 1

i l

1 l

l I

i I

l I I J

l

, . - ~ . - ,

. SNC Response ta NRC RAI Related to Chapt:r 3.4 l g i s

l 5

i

( ' Section 3.4 - Reactor CW System (RCS)  !

ITS 3.4.1

- RCS Pressure. Tu.wieure. and Flow Decarture from NW*, Boiline (DNB) Limits l j

i NRC Question:

I =

f

1. ITS SR 3.4.1.4 adds a NOTE to this surveillance that it is not required until 7 days after i

greater than 90% power but the STS NOTE specifies a delay of only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> aAcr greater  !

~

than 90% power. The time limit speci6ed in ITS SR 3.4.1.4 is seven (7) days. The CTS ~ j does not specify a time for the SR. This is a deviation from both the CTS and the STS and l is, thereforer, Beyond Scope  !

SNC Response:  !

l

'Ihe 7-day allowance for delay of performance of the RCS flow rate is a more conservative requirement than currently exists in the C'IS. In the CTS, no such restnctions apply. 'Ihc only l

l requirement is that the surveillance be performed every 18 months (up to a grace period of 22.5  ;

l.

months). 'Ihe speci6 cation of a power level greater than or equal to 90%, ensures the results of  !

the surveillance are accurate (higher power, more accurate results) and the requirement to j perform the surveillance within a specified time aAer reaching this power level ensures that the  !

RCS flows are verified in a timely manner. Maintaining the current licensing basis of no tune l restriction would be a less conservative option that the proposed subnuttal Based on operating i experience and the precision heat halance being a complicated measurement that requires stable j thermal power conditions, adopting the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit would not ensure that the surveillance  !

was i=fui. .d under the plant conditions which are necessary for this test. Seven days provides adequate time to set up for the measurement, with allowances for typical instrumentation problems, and to achieve stable condations without adversely affecting safety. In addition, the 7 l

~

day limit fori fvi ieg this same surveillance has been previously approved by the NRC for Vogtle (CTS and ITS). Adopting this new time limit for Farley casures' consistency between '

these two SNC plant sites. In the conversion to the ITS, SNC used the standard definition of beyond scope changes. A beyond scope change was defined as a change to a currently existing i requirement in the C'I3 and a change to the STS. As the CTS does not specify a time limit, this is not a change to a limit currently existmg and therefore is not a beyond scope change.'  ;

ITS 3.4.2 RCS Maimum Temperature for Cnticahty NRC Question:

2. - CTS 3.1.1.4 Action requires that with RCS operating loop temperature less than 541 deg F, restore temperature to within limits within 15 minutes or be in HOT STANDBY (MODE 3) in the next 15 minutes. 'Ihis change is consisteat with the STS and isjustified by DOC 4A, however, the change is less restrictive, not administrative. Provide discussion and justafication for the less restrictive change.

Page 1 of16 L _

SNC Response is NRC RAI Related ts Chapt:r 3.4 SNC Response:  ;

5 In the CTS, the total time allowed for restoration of RCS operatmg loop temperature before the l' unit must be in MODE 3 is 30 minutes. In the STS, the total time allowed for restoration of RCS l

operatmg loop temperature before the unit must be in MODE 3 is 30 minutes. In either case, if .

l the ts.giere is restored to above the LCO limit, prior to the expiration of the specified time  :

intervals (Completion Times), the ACTION statement (Required Actions) may be exited per LCO  :

j 3.0.2. In the STS, restoration to operable status is recognized as being an available alternative in i all cases. 'Iherefore, CfS actions are simplified to eliminate the obvious alternative of restoring  !

to operable status (restonng temperature to within limits). As this alternative action is always an ea ~-A option (restoration to meet the LCO is always permissibic) and the total completion time .

is the same for both the C'lli and the STS, this change is not less restrictive. This is an  :

=A=iaWative change made to conform with the format and presentation of this information in l l the STS.  :

NRC Question:  ;

3. CTS 4.1.1.4.b requires RCS temperature to be determined to be greater than or equal to 541  ;

i *F with the reactor critical and T, less that 547 'F with T, -Tw Deviation Alarm in the j alarm condition. While the T,-Tw deviation and low low T, are in brackets, this only l means that you insert or choose your plant specific condition. While DOC 7A pmvides a l justification, the ITS is a deviation from the CTS and S'Ili and is therefore Beyond Scope SNC Response:

I j The T., -Tw Deviation Alarm is calculated using a median T, and a default or low limited

! value for Ta of 547 'F prior to turbine loadmg As Ta is compared to the median signal, one I loop T., could be < 541 *F prior to actuating the T., -Tw Deviation Alarm Use of the Lo-Lo l l T, Alarm provides an alann any time the t-.wsre in any of the three loops is below 543 'F.

In addition, three Trip Status Light Box indications are available to the operator to indicate individual loop T., < 543 'F. 'Ihe T., -Tw Deviation Alarm is currently set to alarm with a deviation ofi 5 'F. However, this setpoint could change in the future. Revising the TS

! requirement to base increased frequency monitoring on the Lo-Lo T., Alarm instead of the T., -

l Tw Deviation Alarm provides greater flexibility for the operator and eliminates concerns over l single failure of the alarm and revisions of the setpoint due to operating variations. While this is a change to the CTS, it is a more conservative limit for the plant. 'Ihis change is consistent with an option of the STS, which is in use at other plants and therefore not a beyond scope change.

l l

l l

Page 2 of16 I

- SNC Response to NRC RAI Related to Chaptsr 3.4 ITS 3.4.3 -

RCS Pressure and Twaure (P/D f.ir=h There are no comments on ITS 3.4.3.

ITS 3.4.4 RCS Imns- MODES I and 2 -

NRC Question:

1

4. De CTS does not specify how many loops you have, it only says all. Dere is no JFD indicating that the "three"is your CTS. Derefore, it is assumed that "three" is the CTS  !

number. 1 1

SNC Response:

! Farley Nuclear Plant is a three-loop Westinghouse PWR. i i

ITS 3.4.5 RCS Loons -MODE 3 l 2

4

[ NRC Question: I

5. ITS SR 3.4.5.2 requires verification of SG secondary side water levels are 2 74% (wide range) for RCS loops. He CTS is 10 % while the STS has a bracketed 17%. His is a  !

Beyond Scope issue. j r

} SNC Response:  !

ne Bases for SR 3.4.5.2 d-==e= the potential impact of not maintaining the minimum level in l the steam generaton. De basis for the level is to ensure that the steam generator tubes remain .

covered, thereby ensuring that the associated loop is capable of providing the heat sink for the i removal of the decay heat. %c CTS Bases do not provide the basis for the 10% steam generator i level. As part of the conversion to the ITS, Farley requested that Westinghouse perform an  ;

evaluation to determine the level =~ - - y to meet the basis stated for SR 3.4.5.2 (as well as SR 3.4.6.2). A bounding level, for the applicable MODES, to ensure that the steam generator tubes -

remain covered was det odoed to be 74% wide range steam generator level. l As stated on the second page of Enclosure 4,'" Bracketed informaten in the STS is confirmed to be applicable to FNP or replaced with information that is applicable to FNP or deleted as

. appropriate. In cases where previously NRC approved CTS information is used to replac- Ge generic STS information in brackets, nojustification for altenng the bracked STS mformation is j provided. De basis for all such changes to the STS is to maintain the current FNP licensing basis >

as specified in the CTS." As this change replaces the bracketed information with information tint  ;

is different from both the current licensing basis and the STS, a JFD should have been included in  ;

Enclosure 5 for this change. It was inadvertently missed duc to the fact that it was replacing l bracketed information. A JFD for this change has been added to the package. While this change l

differs from the current licensing basis, it requires a greater steam generator level based on a i Farley specific evaluation to ensure that the steam generator tubes remain covered, and is  !

therefore a more conservative position for plant operation Fage 3 of16 l

)

SNC Response to NRC RAI Related ta Chapt:r 3.4,

! ITS 3.4.6 l RCS Loons- MODE 4 l NRC Question: s

6. Same as comment 4 above. '

SNC Response:

See response to comment 4.

l ITS 3.4.7 ,

RCS Imos- MODE 5. Imns Filled .i NRC Question:

I

7. ' Same as comment 4 above.  ;

SNC Response: i l See response to comment 4. l

\

l NRC Question: r

8. CTS 3.4.1.4 requires M (RHR) loops to be Operable M DME in operation. 'Ibe ITS, 1 while consistent with the STS, requires One Operable and in operaten and the option or an additional RHR loop QR two secondary side SGs with water level greater than 17%. 'Ihe #

We says three (3) filled loops M at least two SGs may be substituted for one RHR.

DOC 2 does not indicate this. Please providejustificatma for the differences i SNC n- ;  :

'Ihe Applicability ofITS LCO 3.4.7 is " MODE 5 with RCS loops filled." The phrase "Ihree

filled Reactor Coolant Loops"in the # footnote of CTS 3/4.4.1.4 is shown to be ircrpned into l the Applicability ofITS LCO 3.4.7 on the markup of CTS page 3/4 4-4a in Enclosure 1. DOC 2A has boon revised to state the followirg: "Ihe option provided by the # footnote to allow 2 steam generators to be substituted for .me of the required RHR loops when 3 RCS loops are 611ed is ir+4 d directly into tbc LCO (new LCO parts a. and b.) and the Applicability statement.

NRC Question:

l-

9. CTS 3.4.1.4 Action b, same as comment 4 above.

! SNC Response:

i See response to comment 4. }

i i l

i l

Fage 4 of16 l

+. SNC Response ta NRC RAI Related ta Chapter 3.4 NRC Question:

10. DOC 3M provides a discussion of the combined =d~= g ofITS 3.4.7 Condition A and B :

for MODE 5-Loops Filled, which is consistent with the STS, but it does not directlyjustify .

the changes from CTS 3.4.1.4 Action a and Action b to ITS 3.4.7 Conditions A and B as {

indicated in the markup. Provide additional discussion andjustification that directly j justifies the changes from CTS 3.4.1.4 Action a and Action b to ITS 3.4.7 Conditions A and. i B. i l

SNC Response: j i

l CTS action b. is revised to reflect the additional requirement to have the RHR loops operable as j stated in STS Condition B. In addition, the generic term " coolant loop"is changed to "RHR loop" consistent with the STS. Dese changes, combined with the allowance to substitute two i steam generators with adequate level as described in the CTS # footnote for an RHR loop, reflect j the =~~- y changes to convert CIS acten statements a. and b. to STS Conditions A and B. A statement similar to that above has been added to DOC 3M.

ITS 3.4.8 l RCS Loons - MODE 5. Loons Not Filled  ;

i Bere are no comments on 3.4.8.

ITS 3.4.9 l Pressurizer j l

NRC Question:

11. CTS 3.4.4 shows a pressurizer water volume at 868 cubic feet with a correspondag %  ;

indicated level. ITS 3.4.9 adopted the STS. He numbers in your analysis assumption are '

the numbers that should be used. Also note that the 92% shown in the STS is 92% of instrument span. ,

t SNC Response:  !

De Bases for the STS state the following: "Ihe intent of the LCO is to ensure that a steam bubble exists in the pressurizer prior to power operation to minimize the consequences of Fa*i=1 overpressure transients. De presence of a steam bubble is consistent with analytical i assumptions." De intent is not to preserve the safety analysis assumption for pressurizer level at the istnation of an accident. Specific numbers in safety analyses vary depending upon which accident is being analyzed. In some cases, a high level is more 1:d ' iar In some cases, a low level is more limiting. The NRC-approved change to NUREG-1431, TSTF-162, clarifies that the LCO limit requirement is to ensure that operation is within the safety analyses assumption of ensunng that a steam bubble exists in the pressurizer. Setting the TS level at the value - ~ ia'~I with the Pressurizer Water level - High trip setpoint satisfies the requirement to maintain a bubble in the pressurizer and emw is to the basis originally agreed upon during the development of the NUREG. A review of the bases for LCO 3.4.9 was performed to ensure applicability to Farley. He intent of this LCO, as stated in the NUREG, is applicele to Farley and has been ad=*~i Page 5 of16

~ -_. . . _ _ _ _ . _ _ _ , _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ . _ _ . - _ _ . _ - _ _ _ _

.- SNCEL,_ _ _ to NRC RAI Related ta Chapt:r 3.4 l

. 1 ITS 3.4.10 Pressurizer Safety Valves  ;

NRC Question:

12. ITS 3.4.10, DOC IM indicates that the LTOP specification is not applicable until RCS cold )

~ leg t..p.. tore is s 310 *F. ITS 3.4.12 shows that t .w.;4re at 325 *F. Which is  !

correct. Explain the disconnect between the DOC and 3.4.12 and clarify the remaining l discussion on the DOC. l l

SNC Response:  !

Dis error was discovered by SNC during the review of the package submitted on March 12,1998 l while preparmg the clean-typed copy of the ITS. The correct RCS cold leg temperature is 325'F. 1 ne clean-typed copy was submitted to the NRC by SNC letter dated April 24,1998. De second l paragraph of that lettes stated the following: "During review of the submitted package for the  !

creation of the clean-typed copy, some changes were made to correct minor editorial errors and j inconsistencies within the package. Dese changes have not affected the previously submitted j SHEs. He changes are included in A*=4-nt 1. Dese pages should be substituted for the i correspondag pages in the March 12,1998 submittal." A*=4--+ 1 to that letter included the  !

above requested correction, which was reflected in the clean-typed copy. 1 ITS 3.4.11 ,

Pressurizer Power Operated Rehef Valves (PORVs)

NRC Question:-

13. STS 3.4.11 is appbcable in Modes 1,2, and 3, and STS SR 3.4.11.2 requires pe forming a l complete cycle of each PORV cach 18 months. ITS SR 3.4.11.2 is also applicable in Modes 1,2, and 3 but ffS SR 3.4.11.2 is modified to require performing a complete cycle of each PORV during Mode 3 or 4 cach 18 months. JFD #1 discusses this deviation from the S'13 as consistent with CTS 4.4.5.'.a. which is correct, but there is no justification for elimmating performing ITS SR 3.4.11.2 in Modes 1 or 2. Provide discussion and l justification for eli=i--+ia: the cycling of the PORVs in Modes 1 and 2, and provide l

additional discussion as to why the MODE of applicability should not include MODE 4.

SNC Response:

The surveillance is applicaible in MODES 1,2, and 3. De specified frequency for this surveillance in the STS is 18 months. De STS Bases for SR 3.4.11.2 states that the frequency of 18 months is based on a typical refueling cycle and industry WM practice. As such, this surveillance would not be performed in Modes 1 or 2. In the STS, SR 3.0.4 requires all surveilla== to be performed withm their "speci6ed frequency" prior to entering the Mode of Applicability (Mode 3 in this case). Herefore, based on the STS requirements, this surveillance would be pufu....#. prior to entry into MODE 3 from MODE 4 i

i ..

l Page 6 of16 i

-- . .. . - . - . _ . _ - . - - . .. .- - _ ~ _ -

SNC Resporse ts NRC RAI Rel:ted ts Chapt:r 3.4 The modification of the surveillance to state that it must be performed in MODE 3 or 4, consistent with the current licensing basis, does not affect the applicability of the LCO. His change merely limits the performance of this surveillance to sptem conditions where temperatures are high enough to eliminate testing uncertainties and allow for entry into the MODE of Applicability without the surveillance being performed, while ensuring that the surveillance is performed prior to entering MODE ?.. Herefore, addition of MODE 4 to the applicability is not warranted. As stated in the CTS bases (page B 3/4 4-2a, page 76 of the RCS submittal), the PORVs are stroke tested during MODES 3 or 4 with the associated block valves closed in order to eliminate uncertainty introduced by testing the PORVs at lesser sptem temperatun:s than expected during actual operating conditions. A statement similar to that above has been added to the ITS Bases discussion for SR 3.4.11.2.

NRC Question:

14. ITS SR 3.4.11.2 NOTE is not clear. Reword to climinate ambiguity. l l

SNC Response:

ne note and associated discussions have been deleted.

l NRC Question:

15. Does the bracketed information in ITS 3.4.11.3 apply to Farley? If, so, retam ITS as well as retaming CTS.

SNC Response:

He statement of SR 3.4.11.3 in the STS discusses testing of solenoid air control valves and check valves on the air accumulators in the PORV control system. Air accumulators are not used at Farley. Derefore, the wording in the STS is not applicable to Farley. At Farley there are two ,

backup methods to supply control gas pressure to the PORVs. A barbrp air supply path may be  !

used or a nitrogen pressure supoly syW:m from redundant nitregen bottles may be used as a i backup to the normal instrumerit air system for control of the PORVs. Together, these two alternate paths and supplies comprise the backup PORV control system discussed in the CTS j surveillance 4.4.5.1.b. ne wording substituted for the STS wording in SR 3.4.11.3 is taken from i the CTS surveillance and modified to adopt the STS format. Derefore, the wording proposed in the submittal maintains the current licensing and design basis of Farley and should be retained.

ITS 3.4.12 Low Temperature Overpressure Protection (LTOP) System NRC Question:

16. CTS 3.4.10.3 Action a requires that with one RHR relief valve inoperable, restore it to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR perform three altemative actions that are specified by CTS 3.4.10.3 Action a.l.i, and CTS 3.4.10.3 Action a.l.ii, and CTS 3.4.10.3 Action a.l.iii.

ITS 3.4.12 Condition D only requires a Completion Time of 7 days. Dere is no DOC discussion for this. Additionally, there is no justification for deleting the ITS reference to MODE 4. Provide discussion and justification for deleting the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restoration requirements of CTS 3.4.10.3 Action a and including the 7 days. Also provide justification for deleting the reference to MODE 4.

Page 7 of16

. _ . . - - _ _ _ _ . - . . _ _ - . . . - _ - _ _ - _ ~ _ _ . _ _ _ _ _ _ _ _ _ - . _ . .

i SNC Response ta NRC RAI Related t3 Chapter 3.4 i SNC Response: l l

%cre are three completion times associated with Condition D in the FNP ITS. Required actions  !

D.1 and D.2 have 24-hour completion times, and Required Action D.3 has a completion time of 7 l days. In the STS, restoration of a component / system to OPERABLE status is always an unstated  ;

option. Ifit is accomplished prior to the completion of the required actions, the LCO may be i exited. Derefore, it is unnecessary to maintain CE 3/4.4.10.3 action a. as it is always included  !

by implication. If the inoperable valve is not restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ,

CTS actions a.1.i, a.l.ii, and a.l.iii would be applicable. As 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed in the CE to i restore the valve to OPERABLE status, the completion time of required actions D.1 and D.2 (CTS actions a l.i and a.l.ii) is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Required action D.3 has a completion time of 7 days ,

consistent with STS required action E.1 and CTS action a.1.iii.  !

i y CTS 3/4.4.10.3 maintains a 7-day restoration time for metian a.l.iii in Modes 5 and 6 as well as I

! Mode 4. Derefore, the words "in MODE 4" have been deleted from STS Condition E (ITS f

Condition D).  ;

i .

A new DOC has been added to Enclosure 2 for CTS 3/4.4.10.3 similar to that above. JFD #5 has been revised to add a discussion of the deletion of the MODE 4 limitation to ITS Condition D.

i NRC Question: i i

l 17. CTS 3.4.10.3 Action c requires a Special Report for RHR relief or RCS vent valves used to l mitigate a RCS pressure transient. DOC 6A states that this report is moved to STS 5.6.4, the Monthly Operatmg Report Be sure that the Reviewer for Section 5.0 sees the comment.-

SNC Response:

l CTS 3.4.10.3 Action c was included in Enclosure 1 of the Section 5.0 subauttal. Changes are addressed in the Section 5,0 submittal.

NRC Question: 1

18. ITS 'SR 3.4.12.2 requires verifying each accumulator is isolated every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and is added to the surveillance requirements shown on CTS 4.4.10.3.1, page 3/4 4-33. Although DOC 4M provides discussion for adding ITS 3.4.12 Condition A, B, and C which include addag accumulator requirements to the ITS, there is no discussion for adding ITS SR i 3.4.12.2 and establishing the frequency at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ' Add discussion and justification for i ITS SR 3.4.12.2.  !

SNC Response:

CTS 3/4.4.10.3 is revised by the addition of a new surveillance, SR 3.4.12.2, related to new I Actions Conditions B and C consistent with the STS. De proposed surveillance addresses the

, new LCO requirement for the accumulators to be isolated. De proposed surveillance provides raamaamhic requirements for ensunng compliance with the new LCO As such, the new i surveillance is applicable and appropriate to FNP. However, the addition of this surveillance in the FNP TS iq,M a new requirement not previously addressed by the TS. As such, the 4

i I~ Page 8 of16 -

---+4*wm---

SNC Response t2 NRC RAI Rel:ted t2 Chapt:r 3.4 addition of this surveillance is considered a more restrictive change. A new DOC, similar to the discussion above, has been added to Enclosure 2 for Section 3.4.

NRC Question:

19. STS SR 3.4.12.7 requires verifying the associated RHR suction isolation valve is locked open with power removed for each required RHR suction relief valve each 31 days. This surveillance is deleted from the ITS by JFD #10 and is based on performance ofITS SR 3.4.12.3 (STS SR 3.4.12.4) as well as control room indications and administrative controls.

Additionally, ITS SR 3.4.12.3 has the frequency increased in the ITS to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from 12 ,

hours. 'Ihe STS bases for STS SR 3.4.12.2 and STS SR 3.4.12.7 together provide solid l fundamentaljustification for both surveillances. Although the equivalent of STS SR j 3.4.12.7 is not contamed in the CTS, deleting it from the ITS appears to be a presentation  :

preference and is not adequatelyjustified as a deviation from the STS by JFD #10 because l there is no plant speci6c design difference necessitatmg this deletion since the CTS requires suction relief valves. Providejusti6 cation for deleting STS SR 3.4.12.7 based ,

upon plant speci6c design or include STS SR 3.4.12.7 in ITS 3.4.12.

l SNC P==p-- ::

'Ihe operability of the RHR isolation valves and associated interlocks is assured by strict administrative controls. Adding the requirement to lock open and remove power to the valves to the existing requirements would not be a conservative action with respect to a leak in the RHR system. 'Ihe Farley Abnormal Operstmg Procedure (AOP) for RHR System Malfunction directs the operator to isolate the affected RHR train from the RCS in the event of RHR icakage. The RHR loop suction valves are located in containment; two inside the missile barrier and two outside. Should an RHRleak occur, isolation of the leak would be greatly delayed by the necessity of obtaimag keys, dressing out, and entermg conf dament to unlock the valves. In addition, if the leak is in containment the valves may be is the vicinity of the leak, exposing the personnel to unnecessary contamination. Therefun:, M: d on the CTS surveillance, continuous valve status indication for these valves in the control room, existing admmistrative controls for mai- :-% TS required valve positions, the location of these valves, and the requirement to have the capability to cycie these valves to terminate leakage events, the addition of the requirement to lock the valves open and remove power is nonconservative and unnecessary. This maintains the current licensing basis for Farley as reflected in the CTS.

ITS 3.4.13 RCS Operational LEAKAGE NRC Question:

20. CTS 3.4.7.2.e requires that RCS leakage shall be linuted to 31 gpm CONTROLLED i LEAKAGE at a reactor pressure of 2235 i 20 psig. This requirement is moved to ITS 3.5.5 consistent with the STS. Be sure that this specification is reviewed by the Reviewer for Section 3.5.

SNC Response:

CTS 3.4.7.2.e was included in Enclosure 1 of the Section 3.5 submittal. Changes are addressed in the Section 3.5 submittal.

l Page 9 of16

)

.- - . - - . - ---~. --.- . - - ~

! 5 j SNC Response ta NRC RAI Rel:ted t2 Chapter 3.4 j

_ i w

i ITS 3.4.14 RCS Pressure Isolation Valve (PIV) I *abe ~

2 l

I NRC Question:

21. Genera! comment - You can not use the surveillances from the ITS and the Frequencies 1 from the CTS when the surveillances are different than your CTS unless it is identified as a l l Beyond Scope Issue and reviewed separately.  ;

SNC Response: I

! i

~

Where such cases have been identi6ed, SNC has addressed the NRC comments.  !

NRC Question

I

22. CTS 3.4.7.2.f provides the maximum allowable leakage of any RCS PIV shall be "as  !

i specified in CTS Table 3.4-1." Per DOC 2A this requirement is maintained by ITS 3.4.14 1 by substituting the phrase "within limit." 'Ihe actual PIV limits are specified in ITS SR  ;

4 3.4.14.1 by verifying leakage is equivalent to less than or equal to 0.5 gpm per nominal l inch of valve size up to a maximum of 5 gpm and this wording is consistent with the STS. .!

i. CTS 4.4.7.2 contains the same wordmg and Table 3.4-1 gives the speci6c allowable ,

maximum leakage in gpm for each PIV. CTS 4.4.7.2.2 states an upper limit of the maumum allowable leakage per Table 3.4-1 which for some PIVs is 3 gpm. Wre is no ,

discussion in DOC 2A or anyjusti6 cation that explains that these are the same limits.

Provide discussion and justification that the limits in CTS Table 3.4-1 in conjunction with  ;

l C'IS 4.4.7.2.2 requirements are the same as ffS SR 3.4.14.1 or retain the CTS.

l l SNC n=7: .

CTS Table 3.4-1 lists Reactor Coolant System pressure isolation valves with nonunal valve sizes  ;

l I

ranging from 6 inches to 12 inches. Two limits for allowable leakage criteria are listed in CTS

[ 4.4.7.2.2; (1) 0.5 spm per inch ofnominal valve size and, (2) an upper limit of the maximum allowable leakage in Table 3.4-1. For example, using only the first criteria, a six-inch valve l

, would have a limit of 3 gpm, a 10-inch valve would have a limit of 5 spm, and a 12-inch valve r

.would have a limit of 6 spm. W valves listed in Table 3.4-1 with a maximum allowable leakage  !

, of 3 spm are 6-mch valves, consistent with the limit of 0.5 gpm per inch of nominal valve size. ,

'Ihe Table has a maximum limit of 5 spm for all larger valves (10 and 12 inch) In each case the more restrictive of critena 1 or 2 are used for maximum allowable leakage. 'Iberefore, the ITS SR 3.4.14.1 requirement to " verify leakage from each RCS PIV is equivalent to s 0.5 gpm per 1 nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure 2: 2215 psig and $

r 2255 psig"is equivalent to the CTS requirements for all valves listed in Table 3.4-1. A

!. discussion similar to that above has been added to DOC 2A.

l NRC Question:

1 .

j ' 23. STS 3.4.14 Required Action A contair:s a NOTE concerning the requirements for valves i used to isolate one or more leaking PIVs which includes a specific requirement that the

, valve "must have been verified to meet SR 3.4.14.1." In ITS 3.4.14 Required Action A this j iN.L .m is deleted which is a deviation from the STS. JFD #1a discusses this deletion

- but the deletion appears to be a presentation preference that is not sufficientlyjustified as a Page 10 of16 w * + ,.x -w .. c,---4 , . - - ,-

SNC Response is NRC RAI Relat:d ta Chapttr 3.4 r

I STS deviation by JFD #1a because there is no plant specific design difference necessitating this deletion. Providejustification for deleting "have been verified to meet SR 3.4.14.1" ,

from the NOTE in ITS 3.4.14 Required Action A based upon plant specific design or  !

include this requirement in flS 3.4.14 Required Action A.

SNC P=7: :::  ;

f De FNP CTS Actions and the corresponding STS Actions require that the valves used to meet the Action isolate the PIV leakage. Failure to isolate the leakage (mect the requirements of the .

LCO) is failure to meet the Action requirements of Condition A and would result in entry into I Condition B and a plant shutdown. Since in order to meet the isolation Actions, the leakage must  !

be monitored after the isolation valves are closed and verified to be within the LCO limits, the  ;

isolation capability of the valves is verified when they are actually used to meet the Action  !

requirements Compliance with the CTS and ITS Action Requirements are adequate to address i PIV leakage and Ae deletion of the STS requirement to only use valves previously tested per SR 3.4.14.1 is consistent with the FNP current licensing basis.

NRC Question: ]

l l 24. ITS 3.4.14 Condition C, relating to the RHR autoclosure interlock function is aided to the ITS, consistent with the STS, and discussed by DOC 8A. Although this additica to the ITS is sufficientlyjustified by DOC 8A, it is a more restrictive change and not an alministrative change. Change DOC 8A to DOC 8M.

)

SNC Response:

CTS 3/4.4.7.2 is revised by the addhion of STS LCO 3.4.14 Condition C. As the addition of this new Condition in the FNP TS represer.ts a new requirement not previously addresse d by the TS, the addition of this Condition is considered a more restnctive change. DOC 8A has been revised by the addition of wording similar to that above and been changed to DOC SM. i NRC Question:

25. CTS 4.4.7.2.2 Insert M provides four NOTES to ITS SR 3.4.14.1 which are discussed by DOC 13A. Only NOTES 1,2, and 3 are actually shown on the STS SR 3.4.14.1 markup.

NOTE #4 cei@ to CTS 4.4.7.7.2.2 footnote # and allows for an alternate means of determining leakage for ALARA concerns DOC 26LAjustifies moving NOTE 4 to the Bases for ITS SR 3.4.14.1. He CIS markup shows CTS 4.4.7.2.2 footnote # as moved to se Ik.,es per DOC 26 LA, but the CTS 4.4.7.2.2 Insert M still shows NOTE #4. %c STS SR 3.4.14.1 markup does not have a NOTE #4. Correct CTS 4.4.7.2.2 Insert M to remove

! NOTE #4 and revise DOC 13A to address that NOTE #4 is moved to the Bases. 1 SNC Response:

l t Note #4 has boca deleted from Insert M and DOC 13A has been revised to delete reference to

! Note #4. DOC 26LA addresses the movement of the # footnote to CTS surveillance 4.4.7.2.2 to 1 i I the ITS Bases.

i i i  !

i 1 Page 11 of16  !

i l

i ___ __ __ .__ _ ____ _ _ __ . _ __ , . _ . __ _. _ _ , _ _ _ .

r i

C l ,

SNC Response is NRC RAI Relat:d tn Chapt:r 3.4  ;

i NRC Question:

26. STS SR 3.4.14.1 requires leakage testing for PIVs and specifies a frequency that in part  !

[ states,"In accordance with the Inse:vice Testing Program AND Prior to entering MODE i- 2...." ITS SR 3.4.14.1 has deleted this nyirement. Changing the presentation of this l Frequency is a Generic Change because it would apply to any plant. 'Ihere is nothing l l unique about Farley, hrefore, retam the Frequency as stated As is the case throughout j l the STS, time limits are imposed wherever there is a Frequency. JFD #5 provides a reason; '

however, a time limit should be specified. Ifit deviates from the STS, then it can be  !

! reviewed as a Beyond Scope Issue. 'Ihe 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have not posed a hardship for other j plants.  !

SNC Response:

1

! The additional frequencies, "In accordance with the Inservice Testing Program AND Prior to

~ I

entering MODE 2...." are not included in the Farley CTS. Stating that the surveillance is  ;

j perfonned in accordance with the Inservice Testing Program is redundant to the 18-month l frequency (as described in the bases). & additional frequency related to Mode 5 entry is an

unnecessary restncton not currentlyimposed on Farley. & deletion of the frequencies is not a j generic change to the STS, but a plant-specific change made to conform to the Farley current

licensing basis.

5 i As described in JFD #5, testmg of the PIVs locally requires that certain plant conditions be i achieved and test equipment be set up to perform the test. Under certain plant conditions, aAer 4 initiating flow through valves,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would not be enough time to establish these test i conditions, set up the test equipment, and perform the test locally. Site personnel responsible for performing the testing do not provide round-the-clock on-site coverage and may need to be called j in to perform the test. PA s of this test on an exigent basis would require additional j planning and risk analysis. Farley's design does not incorporate any remote testing systems for 4

such tests. hrefore, deleting the 24-hour time limit revises the STS consistent with the Farley design and current licensing basis.

1 Not all frequencies have a specific time limit imposed Some frequencies are based on events or conditions. SR 3.1.3.2 has a frequency of"Once each cycle." SR 3.1.3.1 has a frequency of i "Once prior to entering MODE I aAer each refueling." SR 3.1.4.3 has a frequency of" Prior to j reactor criticality after each removal of the reactor head." SR 3.1.4.3 has a frequency of" Prior to j initiation of PHYSICS TESTS." SR 3.3.2.10 has a frequency of" Prior to reactor startup."

! Therefore, a frequency of"Following valve actuation due to automatic or manual action of flow through the valve"is acceptable and consistent with the Farley current licensing basis.

NRC Question:

27. STS SR 3.4.14.2 requires verifying the open permissive feature of the RHR isolation valve autoclosure interlock function and is deleted from the ITS. CTS 4.5.2.d.1 only verifies the autoclosure feature of this interlock which is consistent with STS SR 3.4.14.3 and is reinM in the ITS as ITS SR 3.4.14.2. Deleting STS 3.4.14.2 is a deviation from the STS.

JFD #2 %== tids deletion, but the deletion appears to be a presentation preference that is not sufficientlyjustified as a S'I3 deviation from the STS by JFD #2 because there is no Page 12 of16

+. SNC Response t2 NRC RAI Rel:ted t3 Chipt:r 3.4 plant specific design difference necessitating this deletion. Providejustification for deleting STS SR 3.4.14.2 based upon plant specific design or include STS SR 3.4.14.2 in ITS 3.4.14 surveillance requirements SNCP= r :

Farley will adopt the STS requirement to verify the open permissive interlock feature for the RHR isolation valves. However, in reviewing this surveillance requirement and the associated acten statements of the LCO, a plant specific issue was discovered which applies to either of the interlock functions being inoperable. At Farley, the location of the electncal switchgear containing the breakers for the RHR isolation valves is subject to very high dose rates in the event of a small break LOCA. Herefore, complying with the Required Action of Condition C of LCO 3.4.14 would place the plant in a condition where, should a small break LOCA occur, the plant could not be placed on normal RHR without --Mbly high exposures to plant personnel. To address the issue of dose during a small break LOCA, the Required Action of Condition C has been revised to require isolation of the valves under administrative controls from the control room ,

' to allow establishment of RHR operaten, should it be required, without unacceptable dose to 1 plant personnel in the event of a small break LOCA.

NRC Question:

l 28 STS SR 3.4.14.3 (ITS 3.4.14.2) contams a NOTE stating "not required to be met when the '

RHR system autoclosure interlock is disabled in acc d.i.ce with ITS SR 3.4.12.7." In ITS SR 3.4.14.2 the phrase "autoclosure interlock is disabled" is replaced with " valves are required open." His change is a deviation from the STS with discussion andjustification .

provided by JFD #4. He JFD #4 justification is based upon changes made to STS 3.4.12.7 (ITS SR 3.4.12.3) and have comment #19 ===arintad with this change to SE SR 3.4.12.7.

He resolution of 3.4.12.7 should have no bearing upon the wording of this NOTE.

Changes to the wordmg would be generic and should be handled as such. Retain STS wordmg SNC Response:-

De wording was changed in this note due to the fact that STS SR 3.4.12.7 was deleted. STS SR 3.4.12.7 requires that operator power be removed for each required RHR suction relief valve, thus effectively disabling the autoclosure interlock. He note in the ITS is revised to reference ITS SR 3.4.12.3 (STS SR 3.4.12.4), wisch .soes not disable the interlock, but verifies the position of the valves. Therefore, since the autouosure interlock was not disabled, the STS note was incorrect as originally worded and was revised to reflect the FNP current licensing basis requirements.

- NRC Question:

29. CTS 4.4.7.2.1.c is a surveillance requirement for CONTROLLED LEAKAGE to the RCP seals. DOC 10A states that this CTS surveillance requirement is moved to ITS LCO 3.5.5 in accordance with the STS, Be sure that this speci6 cation is reviewed by the Reviewer for Section 3.5.

. Page 13 of16 l

I

l i

SNC Response to NRC RAI Related to Chaptsr 3.4 -

, l l

j SNC Response.

)

t '

CTS 4.4.7.2.1.c was included in Enclosure 1 of the Section 3.5 submittal. Changes are addressed i . in the Section 3.5 submittal.- l l

! ITS 3.4.15 i 1- RCS Imkage Detection Instrumentation l i  !

l NRC Question: l l  !

l 30. LCO 3.4.15 differs from both the CTS and SU and should be reviewed as a Beyond Scope I

j. Item. Some features of the STS have been modi 6ed as well as some in the CTS. I I

SNC Response:

i & CTS required RCS leak detection systems have been incorporated into the STS format. The 1 Farley design of the cantainment sump precludes it from being quali6ed as a leak detection L system per Regulatory Guide 1.45. Actions related to the cantninment sump have been deleted

from the ITS and the STS modified accordingly. & CTS has becen marked up to reflect the i adoption of additional STS requirements for the existire Farley systems. Changes made to the STS are only made to reflect the required systems from the CTS. As such, this is not a Beyond j_ Scope change.

j

ITS 3.4.16 ECS Snecific Activity i NRC Question:

i j 31. ITS SR 3.4.16.3 adds a NOTE to the sample requirement for determination of E bar stating

the sample is not required to be p.La ;;d until 31 days after the other specified conditions are met. DOC llM discusses andjustifies this change as more restrictive but the change appears to be a less restrictive change to CTS Table 4.4-4 item 3* Table Notations hec ==

the ITS SR 3.4.16.3 NOTE extends the sample interval an extra 31 days beyond the 6 l

month speci6ed sample interval including the requirement of"after 2 EFPD and 20 days of j Power Operation " Although the change is consistent with the STS and acceptable, is it more orless restrictive?

I SNC Response:  ;

I In the CTS, the item 3

• table notation does not state a maximum time, after the muumum required time has elapsed, to perform the surveillance. 'Ihe item 3

• table notation is intended to ensure that the radioactive materials are at equilibrium so that the analysis for E bar is representative and not skewed by a crud burst or other similar abnormal event. This note modifies the 6-month surveillance frequency such that the surveillance isn't due at 6 months if the ==winted requirements of the note are not met. The maximum time is controlled by the sample frequency listed in Table 4.4-4, with the associated footnote, and surveillance requirement 4.0.2. Surveillance requirement 4.0.2 states that each surveillance requirement shall be

. performed within the speci6ed time interval, with a maximum allowable extension not to exceed 25% of the surveillance frequency.'. If the power history of the plant were such that the Page 14 of16

SNC Response to NRC RAI Related to Chaptsr 3.4 requirements of the

• footnote were met I day after the 6 month suncillance frequency was exceeded, the surveillance could be performed any time in the next 44 days (balance of the 25%

extension past the 6 month frequency) as needed to facilitate surveillance scheduling. However, with the note adopted from the STS, the surveillance would have to be performed within the next 31 days as stated in the Bases Therefon:, this is a more restrictive change.

NRC Question:

32. CTS Table 4.4-4 item 4 requires Isotopic Analysis for Iodine including I-131, I-133, and I-135. The sample and analysis frequency for this analysis, Table 4.4-4 Item 4.a and Table ,

4.4-4 Table 4.b, including the requirements of Footnote #, have been included in ITS 3.4.16 ,

Required Action A.1, B.1 and ITS SR 3.4.16.2 which sample for DOSE EQUIVALENT I-  !

131 specific activity. The Table 4.4-4 Item 4 sample type (Isotopic Analysis for Iodine l including I-131, I-133, and I-135) is deleted and does not appear in ITS 3.4.16. There is no discussion and justification for this deletion. Here is no discussion that Table 4.4-4, Item 4, (Isotopic Analysis for Iodine including I-131, I-133, I-135) is equivalent to ITS 3.4.16 sampling requirements for DOSE EQUIVALENT I-131 specific activity or that only  ;

sampling for DOSE EQUIVALENT I-131 specific activity satisfies the requirements of '

CTS Table 4.4-4 item 4. Provide discussion and justification for deleting Table 4.4-4 Item j 4, Isotopic Analysis for Iodine including I-131, I-133, I-135, such that sampling for DOSE  !

EQUIVALENT I-131 specific activity per ITS 3.4.16 is sufficient. I SNC Response:

DOSE EQUIVALENT I-131 is defined in the TS as that concentration ofI-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture ofI-131, I-132, I-133, I-134, and I-135 present. Therefore, performing an analysis for DOSE EQUIVALENT I-131 includes I-131, I-133, and I-135. He suneillance performed to comply with the requirements of Table 4.4-4 Item 4.a and 4.b of the CTS includes analysis of all 5 Iodine isotopes. Performance of SR 3.4.16.2 will also include analysis ofI-132 and I-134, in I addition to the three isotopes listed in Table 4.4-4 Item 4. As such, performance of SR 3.4.16.2 I' will envelope the requirements of Table 4.4-4 Item 4 and is slightly more restrictive. A DOC has been added to the package with a discussion similar to that above.

CTS 3.4.6  ;

Steam Generators. No exolicit ITS section l NRC Question:

33. CTS 4.4.6 Steam Generator Surveillance requirements are moved to the Steam Generator Tube Surveillance Program of the Administrative Controls section ofITS 5.5.9. Be sure that this specification is reviewed by the Reviewer for Section 5.0.

SNC Response:

CTS 4.4.6 was included in Enclosure 1 of the Section 5.0 submittal. Changes are addressed in the Section 5.0 submittal.

Page 15 of16

i SNC Response to NRC RAI Related t2 Chapt:r 3.4 l 4

CTS 3.4.8

{

Chemistry. No explicit ITS section There are no CTS 3.4.8 comments.

CTS 3.4.10.2 Pressurizer. no exoticit ITS section.

{

There are no CTS 3.4.10.2 comments.

CTS 3.4.11 Structural Inteerity. No exoticit ITS sectim NRC Question:

l

34. CTS 3.4.1I specifies requirements for maintaining structural integrity of ASME Code Class  !

components. Per DOC ILA and DOC 2A, all test and inspection requirements specified in i CTS 3.4.11 are either maintained in ITS 5.5.7 and 5.5.16 or are required by regulation and covered by the IST program ofITS 5.5.8. Be sure that this specification is reviewed by the i Reviewer for Section 5.0.

SNC Response:

Insert K in Enclosure 1 of the Section 5.0 submittal includes the wording from CTS 4.4.11.2, as modified and discussed in the Section 3.4 submittal, and CTS 4.4.11.3. He DOCS in Enclosure 2 of the Section 5.0 submittal reference the CTS basis for each of these new ITS programs (ITS 5.5.7 and ITS 5.5.16). Insert K in Enclosure 1 of the Section 5.0 submittal also includes the STS wording for the Inservice Testing Program. De associated DOC in Enclosure 2 of the Section i 5.0 submittal references CTS 4.0.5 as the basis for this new ITS program (ITS 5.5.8) and states that differences are discussed in the Section 3.0 submittal.

CTS 3.4.12 Reactor Vessel Head Vents. No exnlicit ITS section NRC Question:

35. There are no CTS 3.4.12 comments. Be sure that this specification is reviewed by the Reviewer for Section 5.0.

SNC Response:

CTS 3/4.4.12 was relocated to the Technical Requirements Manual as discussed in DOC 1R for CTS 3/4.4.12 because it did not meet any of the 4 criteria of 10 CFR 50.36. It was not included in the Section 5.0 submittal.

Page 16 of16