ML20236X254

Forwards Addl Info Re TS Conversion Application for Sections 3.1,3.2,3.5,3.9 & 4.0 as Requested by NRC in s. Marked-up TS Pages Included

ML20236X254
Person / Time
Site:	Comanche Peak
Issue date:	08/05/1998
From:	Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-98182, NUDOCS 9808070232
Download: ML20236X254 (300)

Text

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"""" 7 3 5 Log # TXX-98182 File # 10010 1UELECTRIC 916 Ref. # 10 CFR 50.90 10 CFR 50.36 c.1mce nriy Senior Vice President

& PrincipalNuclear Oficer Aygugt9,3g9g U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 TECHNICAL SPECIFICATION CONVERSION APPLICATION ADDITIONAL INFORMATION FOR SECTIONS 3.1,3.2,3.5,3.9, AND 4.0 REF: 1) TU Electric letter, logged TXX-97105, from C. L. Terry to the NRC dated May 15,1997

2) NRC letter from T. A. Polich to C. L. Terry dated June 17,1998, conceming sections 3.1 and 3.2

3) NRC letter from T. A. Polich to C. L. Terry dated June 17,1998, concerning sections 3.5,3.9 and 4.0 Gentlemen:

TU Electric requested an amendment to the CPSES Unit 1 facility operating license (NPF-87) and Unit 2 facility operating license (NPF-89) by incorporating changes to the CPSES Units 1 and 2 Technical Specifications (TS) as provided in reference 1. The NRC staff requested additional information regarding Section 3.1, " Reactivity Control Systems," Section 3.2, " Power

' Distribution Systems," Section 3.5, " Emergency Core Cooling Systems," Section 3.9,

" Refueling Operations," and Section 4.0, " Design Features" of the proposed TS changes in references 2 and 3. The requested information is provided in the attachments to this letter as are any additional changes needed for Sections 3.1,3.2,3.5,3.9, and 4.0 as identified by TU Electric.

}

This letter ar.d the attachments are not a supplement to reference 1. A supplcr.r' to reference 1 will be provided at a later date. Any deviations from the responses provided in ;hk

[ letter will be discussed in the supplement.

l The only commitment contained in this letter is to provide a supplement to reference 1 at a later date.

Y;1 9800070232 990905*

PDR P

ADOCK 05000445 PM 9o

^Vvv.) . 'r s

COMANCllE PEAK STEAM ELrCTRIC STATION P.O. Box 1002 Glen Rose, Tesas 76043-1002 l

u____

TXX-98182 Page 2 of 2 l

if you have any questions concerning the content of this letter, please contact Mr. Bob Dacko l (254-897-0122).

I Sincerely,  !

C. L. Terry By:

I D. R. Woc dlan Docket Li':ensing Manager l BSD/bd Attachments 1. Section 3.1, " Reactivity Control Systems"

2. Section 3.2, " Power Distribution Systems"

3. Section 3.5," Emergency Core Cooling Systems"

4. Section 3.9," Refueling Operations"

5. Section 4.0,

• Design Features.'

c- E. W. Merschoff, Region IV J. l. Tapia, Region IV Resident inspectors, CPSES T - 1. Polich, NRR

' Mr. Arthur C. Tate i Bureau of Radiation Control Texas Department of Public Health .

l 1100 West 49th Street Austin, Texas 78704 i

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- - _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ _ - _ _ _ _ _ _ _ _ - _ _ . _ _ - _ _ _ _ _ _ to TXX-98182 Page 1 of 5 JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS CTS 3/4.1 - REACTIVITY CONTROL SYSTEMS ITS 3.1 - REACTIVITY CONTROL SYSTEMS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES i

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Attachment 1 to TXX-98182 Page 2 of 5 INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER i Q3.1.G-1 DC, CP, WC, CA YES 3.1-1 WC, CA NA 3.1-2 DC,CP YES 3.1-3 DC, CP, WC, CA YES 3.1-4 DC,CP YES 3.1-5 CP YES 3.1 -6 DC,CP YES 3.1-7 DC,CP YES l 3.1-8 DC,CP YES 3.1-9 DC NA 3.1-10 DC,CP YES 3.1-11 DC,CP YES 3.1-12 DC,CP YES 3.1-13 DC, CP, WC, CA YES 3.1-14 WC NA 3.1-15 DC, CP, WC, CA YES 3.1-16 DC, CP, WC, CA YES 3.1-17 CP YES 3.1-18 CP YES 3.1-19 WC, CA NA 3.1 20 DC,CP YES 3.1-21 DC,CP YES 3.1-22 DC NA 3.1-23 WC NA 3.1-24 DC, CP, WC, CA YES 3.1-25 DC, CP, WC, CA YES 3.1-26 CP YES 3.1-27 DC, CP, WC, CA YES 3.1-28 DC, CP, WC, CA YES CA 3.1-001 WC, CA NA CA 3.1-003 CA NA CA 3.1-004 CA NA l

l CP 3.1-ED CP YES CP 3.1-002 CP YES CP 3.1-003 CP YES DC ALL-001 (3.1 changes only) DC NA DC ALL-002 (3.1 changes only) DC NA

Attachment 1 to TXX-98182 Page 3 of 5 INDEX OF ADDITIONAL INFORMATION (cont.)

ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER DC 3.1-ED DC NA DC 3.1-001 DC,CP YES TR 3.1-001 DC, CP, WC, CA YES TR 3.1-003 DC, CP, WC, CA YES TR 3.1-004 DC, CP, WC, CA YES TR 3.1005 DC, CP, WC, CA YES TR 3.1-006 DC, CP, WC, CA YES WC 3.1-ED WC NA

Attachment 1 to TXX-98182 Page 4 of 5 JOINT LICENSING SUBCOMMITTLd METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:

1. Each licensee is submitting a separate response for each section.

2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), *NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

3. If a licensee initiated change does not apply, "NA" has bees entered in the index column labeled " ENCLOSED" and no information is provided in the iesponse for that licensee.

4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,3B, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as " PLANT SPECIFIC RESPONSE ......."

5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.

, 6. A marginal note (the Additional Information Number from the index)is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.

7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request l to reflect the changes being made by one or more of the other licensees. These l

changes are not included in the additional information for the licensee to which the

change does not apply, as the changes are only for consistency, do not technically l

affect the request for that licensee, and are being provided in the additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment request supplement to be provided later.

Attachment 1 to TXX-98182 Page 5 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued)

8. The item numbers are formatted as follows: (Source][lTS Section]-[nnn]

Source = Q - NRC Question CA - AmerenUE DC-PG&E WC - WCNOC CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL" is used for the section number.

nnn = a three digit sequential number or ED (ED indicates editorial correction with no impact on meaning) 1 i

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f ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q3.1.G-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.1.x Bases '

General There have L'en a number of instances that the specific changes to the STS Bases are not properly identified with redline or strikeout marks.

Comment: Perform an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal.

FLOG response: The submitted ITS Bases markups for Section 3.1 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of bra::kets and reviewer's notes). Most of the differences were editorial in nature and would not have affected the review. Examples of editorial changes are:

1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced.

2) Changing a verb from singular to plural by adding an "s" without " redlining" the "s".

3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g.,

SR3.6.6A.7).

4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference.

5) In some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.

6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.

7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was '

determlned to not be applicable, the text was then struck-out and remains in the ITS l Bases mark-up.  !

Differences of the above editorial nature will not be provided as attachments to this response. The pages requiring changes that are more than editorial and are not consistent

with the markup methodology are attached.

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ATTACHED PAGES:

Enci5B B 3.1-47 Title B.1. B.2. and B.3 should have been redlined B 3.1-48 Revised titles incorrectly struckout/ redlined B 3.1-49 Revised title incorrectly struckout/ redlined i

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1 Rod Position Indication B 3-h8 3;n7 BASES full power operation since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

L2 Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3 2).

The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above. I I

Bilr B12Fand Whenimoreitharone; DRRIIperj group ; failfadditjonallectjons are; pecessaryitoiensure'o that:t acceptable;poweredistribution:11 sits areLoaintained;InjnimuelSDE1sinaintained;iand;the; potential effectstoftrod;misalignmentionfassociatedl accident analyses;am limited;;ghesindirect position; determination:available;via moyablelincoteldetectorsM11LeinimizeithelpotentiaEforired Eisa11gnment!

Thelpositioniof thel rods,;may;beideterminedlindirectlyiby;usefot thef..movableiincore; detectorsn1Thef Required LAction;mayZal so:: be sati sfied;byl ensuting;at Eleast;once: pet:8ThoursithatJ,5satjsfies LC013.'2; g F 4: sat 1sfies!LC0j3;2i2,EandiS WTDO W; MARG 1 01szwithin

theHiej ts;provjded

in ithe;COLR;: provided!the;;nonindicating trods have. not:been;novedMVerification Lof;RCCC position;once; per 8;hoursd stadequate;forfal)owingLcontinued2fu11 power; operation i forfa 11mi.tedlj241hout period asincejthefprobabiljtyso f simul taneous]L having;airod? signi ficantly_outioC positionsandici eventisensitive tolthatirod.' position [1sismal M The124;hout Completion 1 Time providespufficientl.tineitoltroubleshootland restoreithe 'ORRIJ systenito; operation;whileiavoiding;thelplant challengesfas.sociated Withia; shutdown]withoutll full: rod.; position

,indicatjonl(RefC:4)]

(continued)

CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-47 72 9/98 L__________________ _--_

Rod Position Indication B 3-1-0 3;12 BASES

. ACTIONS BTit?B:2"andB3EJcontinued)

Based on; operating; experience;1normalipower; operation;does:not requ.ireiexcessiveirod movement;L;If;cnejprimore;rodsihasi been significantlyxmoved,'%theLRequired1Actionsfici1[orlC:2;belowlis required;

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2 o.3,3,s.1 OGB1and V i These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last l determined, the Required Actions of A.1 and A.2;ot BJ1 are still i

appropriate but must be initiated promptly under Required Action C.1 to beginLjndirectly verifying that these rods are still properly positioned, relative to their group positions usi ng ; t.helsovabl e;inciY detectors .

If, within E4B hours, the rod positions have not been determined. THERMAL POWER must be reduced to s 50% RTP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of E4B hours provides an acceptable period of time to verify the rod positions.

A' ~ 0 3.1.G 1 CDG.1.1 and{pG.'.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the DRPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means (e.g;~, observation:of appropriate:DRP R status 11ndications)athat the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are s 12 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate.

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0 3.1.G 1 Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provides an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to s 50% RTP.

(continued)

CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-48 7/29/98

Rod Fosition Indication B 3-1-6 3ild BASES 6

ACTIONS o.3.1.G.1 (continued)

If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a H0DE in which the requirement aoes not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, for reaching the required H0DE from full power conditions in an orderly manner and without challenging plant systems.

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t SURVEILLANCE SR 3-1-8 3'.177.1 l REQUIREMENTS Verification that the DRPI agrees with the demand position within i fle}12 steps ensures that the DRPI is operating correctly.s Verification;atl.24 E480 120; and1228Jstepsifor;the; control' banks and at:18,1210 Eand'2281 steps;forithedshutdown; banks: provide #

assurance'thatithelDRPI;i.s; operating correctly _overhthe) full rangeLoff indication? Since the DRPI does not display the actual shutdown rod positions between 18 and 210 steps, only points within the indicated ranges are required in comparison.

The [IS month] ricquency is based on the accd to pcrform this Surveillance under the conditions that apply during a-pht cutage and the pctcatial for unaccessary plant transients if the SR werc pcrforacd with the rccctor et pcwcr. Operating expericace he; shown thcsc componcats usually pass the SR when pcrforacd at a Frcqucacy of cacc cvcry [IS acnths]10Ath;. l Thereforc, the Frcqucacy was concluded to be acceptable fica a reliability standpcint.ThisLsurveillance?1s7 performed prior 1t o reactorferiticality;afteteach removalfofithe reactordvessel z head,lsinceithere;is,. potential; for; unnecessary plant transjents s if;the;SR wereLperformed with the r.eactorlattpoweri REFERENCES 1. 10 CFR 50, Appendix A, GDC 13. 1 l

2. FSAR, Chapter E153

3. ISAR, Chaptcr [1S].

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1 CPSESMark-up ofNUREG-1431 Bases - 1TS 3.1 B 3.1-49 7/29/98 i

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-2 APPLICABILITY: CP, DC REQUEST: 3.1.1 Shutdown Margin (SDM)(Comanche Peak and Diablo Canyon)

DOC 01-06-A CTS 3/4.1.1 Applicability ITS 3.1.1 Applicability Comment: According to the Conversion Comparison Table," MODE 2 with Keff < 1.0" and " MODE 5" are added to the Applicability section of TS 3.1.1 for Comanche Peak and Diablo Canyon. All of the FLOG ITS Sections 3.1.1 have these applicability requirements included in the ITS and not in the CTS. Provide a discussion for Comanche Peak and Diablo Canyon explaining / justifying these changes.

FLOG Response: In the CTS, shutdown margin is controlled via LCO 3.1.1.1 for MODES 1,2, 3 & 4 and LCO 3.1.1.2 for MODE 5. Rod insertion limits are controlled by LCO 3.1.3.5 for shutdown rods, MODES 1 and 2, and LCO 3.1.3.6 for control rods, MODES 1 and 2. When the reactor is critical (MODE 1 and MODE 2 with Keff 21), shutdown margin is assured, via the CTS, by assuring that rod insertion limits are met (see SR 4.1.1.1.16.) When the reactor is not critical (MODE 2 with Keff < 1, MODE 3, MODE 4 and MODE 5) shutdown margin is assured oy assessing boron concentration, temperature, etc. (see SR 4.1.1.1.1e and 4.1.1.2). The CTS requirements have been clarified and reorganized such that the rod insertion limits and shut margin requirements for MODES 1 and 2 with Keff 21 from CTS 3.1.1.1,3.1.3.5 and 3.1.3.6 are converted to ITS 3.1.5 and 3.1.6. The shutdown margin requirements for MODE 2 with Keff < 1, MODE 3, MODE 4 and MODE 5 from CTS 3.1.1.1 and 3.1.1.2 are converted to ITS 3.1.1. This reorganization does not change requirements but presents them is a more logical manner and therefore is a purely administrative change.

DOC 1-06-A will be revised to add the following:

In the ITS format, the SHUTDOWN MARGIN in MODE 1 and MODE 2 with ken 21.0 is controlled through compliance with control rod insertion limits (ITS LCO 3.1.5 and LCO 3.1.6). For those modes or conditions in which compliance with control rod insertions limits is not required, the SHUTDOWN MARGIN is verified in the more traditional manner by consideration of such factors as Reactor Coolant System boron concentration, coolant temperature, xenon and samarium concentrations, etc. Thus, the applicability of CTS 3.1.1.1 is modified by this change to be applicable to MODE 2 with k,, < 1.0 as well as the current MODES 3 and 4. This change is considered to be administrative in nature, because, when the reactor was critical ( Mode 1 and 2 ,

with K,,21), the SHUTDOWN MARGIN was determined, in accordance with SR 4.1.1.1.1.b, by J l verifying comphance with the control rod insertion limits.

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in addition, the SHUTDOWN MARGIN requirement, surveillance, and actions are the same for operation in MODE 5 as for operation in MODES 3 and 4. Therefore, the specifications have been combined to include MODE 5 with MODES 3 and 4. The change is considered to be administrative in nature, because there is no change in the LCO, ACTIONS, or SURVEILLANCE j REQUIREMENTS. See also Change 2-01-A.

ATTACHED PAGES:

Enci 3A 2 .

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CHANGE NUMBER HSE DESEPTION (1) Deletes an inappropriate Applicability (i.e., the surveillance (s) should not apply in those modes when the rods are not required to be Operable): and (2) Deletes redundant requirements (i.e., the requirements are properly and fully addressed in the specifications related to rod alignment / operability and insertion limits). l

[The Mode 1 and 2 requirements in the CTS specification for SDM is acceptable because they are redundant to the requirements in the rod control specifications.]

01 05 LG The list of specific items to be considered in the performance of a Shutdown Margin verification is removed to the ITS 3.1.1 Bases. This change is of the type that moves unnecessary details from the specifications while leaving the overall safety requirement intact.

01 06 A This change revises the SDM LC0 Applicability to MODE 2 with k,,, < 1.0 MODE 3, and MODE 4. This change also creates a new Core Reactivity LC0 based on 0 3.1-2 ITS 3.1.2. This is consistent with NUREG 1431. InJthe ITS: format; therSHUTDOWN . MARGIN;jn MODE 111and MODE l2 with k,,,T1.0:1s: controlled;throughicompl1anceiwith controtrodiin.sertionjlimits:(ITS:LC0:3.1.51andlLC0 3;1;6)Eforithoselmodeslorl.gonditionsyjn which compliance with;-contro11rodjinsertions limitsiis;notfequiredFthe SHUTDOWN' MARGIN isiverifiedfin'.thelmore traditic9allmanner by considerationLof_;such:; factors;as R.eactor;c Coolant System boronfconcentration;;coolantEtemperaturel:xenomand samarium concentrationsJetc. 2Thus',";the;applicabilitylof CTSL311,1;;1Jsimodified by this: change;toibe; applicable to MODE?2fwjthl:k,,rXI;0iasLwellf as;the current; MODES 3'and 47;This changelis; considered to(be administrative,in natured because; when1the: r9 actor'was;criticaE(i Mode 11and 2)ithl K,r/E1)L the ; SHUTDOWN l MARGI[wa s 1determi ned, fin accordance;withtSR 4;111'.1:1:bnby; verifying; compliance with:the contro1Jrodfinsertion[11mits.?

InfadditionC the SHUTDOWN MARGINirequirement; {

surveillance;" andlactions;are;the;samef forloperation in MODE;5:as7for; operation?in MODES:3;and;4 C Therefore, the specif1 cations;have;been combined;to: include ~ MODE;5;with MODES;3 and;4 G TheLchange11s considered to be l administrative jn; nature,7because there;is no change..in  ;

the;LCO,; ACTIONS,7or SURVEILLANCE REQUIREMENTS 1See also Change 2iO1M;j l

l CPSES Description of Changes - CTS 3N.1 2 7/29/98 j l

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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-3 APPLICABILITY: CA, CP, DC, WC REQUEST: 3.1.1 Shutdown Margin (SDM)(All FLOG Plants)

DOC 01-10-M CTS SR 4.1.1.1.1 ITS SR 3.1.1.1 Comment: The justification for modifying applicability of SR 3.1.1.1 is inadequate; it only refers to consistency with NUREG-1431. Also, it is not apparent why this change is not applicable to Wolf Creek and Callaway.

FLOG Response: For DCPP and CPSES, DOC 01-10-M is revised to state the following: "In the ITS format, the SHUTDOWN MARGIN in MODE 1 and MODE 2 with keff 21.0 is controlled through compliance with control rod insertion limits. For those modes or conditions in which l compliance with control rod insertions limits is not required, the SHUTDOWN MARGIN is verified in the more traditional manner by consideration of such factors as Reactor Coolant System boron concentration, coolant temperature, xenon and samarium concentrations, etc.

Thus, the applicability of CTS SR 4.1.1.1.1.e is modified by this change to be applicable to j MODE 2 with keff < 1.0 as well as the current MODES 3 and 4. This change is more restrictive, I

in that CTS 4.1.1.1.1.b addresses MODES 1 and 2 with keff 21.0, and CTS 4.1.1.1.1.e addresses MODES 3 and 4. MODE 2 with keff < 1.0 is not specifically addressed in the CTS.

See also revised Change 01-06-A, which provides a broad discussion of how the applicabilities for CTS 3.1.1.1,3.1.1.2,3.1.3.5 and 3.1.3.6 have been revised."

The Wolf Creek and Callaway Technical Specifications were modified by Amendment 89 and 103 respectively, to contain a MODE 3,4, and 5 Specifications for " Shutdown Margin" and a separate MODE 1 and 2 Specification for" Core Reactivity." This eliminated the need for individual MODE applications under the Surveillance Section. Wolf Creek and Callaway used DOC 01-02-M to make the MODE 2 with Keff<1.0 change to both the LCO and the SR. This makes DOC 01-10-M not applicable to Wolf Creek and Callaway (see Enclosure 3B).

ATTACHED PAGES:

Enci 3A 2a

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CHANGE l NUMBER NSBC DESCRIPTION l

01 07 LS 16 The term "immediately" is changed to "15 minutes" which is '

more specific. The term "immediately" simply specifies prompt action. The completion time of 15 minutes is meant to clearly state a completed action. The requirements of this action are met only if boron is already being injected at 15 minutes. This time period provides

adequate time for the operator to align and start the l required systems. This is consistent with NUREG 1431.

01 08 A The technical contents of this surveillance requirement l (verification of shutdown margin through compliance with

rod insertion limits) in Mode 1 and Mode 2 with k,,, 21.0 have been incorporated into LC0 3.1.6 of the improved TS. i 01 09 A The SR for verification of the estimated critical condition during the approach to criticality is moved to ,

ITS SR 3.1.6.1. '

l 01 10 M CTS SR 4.1.1.1.le is modified by this change to be applicable to MODE 2 with k,,, < 1.0 as well as the current MODES 3 and 4. This is consistent with NUREG- 0 3.1 3 1431. In theJITS u format;;thei UTDOWNiMARGINiin;MODEll SM and; MODE 12yith keff21;0jjs! controlled 1through 1 l compl iance; with lcontro11 rod;insertjotl imits.E For; those l modes ' ori conditions 11n Lwhich incompliance _with;controE rod

! insertions 111mitsitsinot1 required 2 the LSMUTDOWNiMARGIN ;i s j verified j inithelmore itraditjonalimanneriby; consideration ofisuch factorslas; Reactor; Coolant 1 System boron concentration;1 cool ant 6 temperature;T xenon;and; samarium concentrationsZetc;1ThusEthell applicability;of1 CTS SR 47111;E1.e.;1simodified;byLthis change ltolbe; applicable;to MODET2;with?x e ffME0;as(well[asithe current; MODES;3 and 4 GThi s: change; i s; nore; restrj etiveEin tthat; CTS 4~.L171;1'.b: addresses; MODES :11and 2;with1ketf 21._0Uand l CTS 14;E1;1;Ee; addresses MODES;3"andLMODE72;with keyf l

5.1.03s not specifically;addressedkjnIthe; CTS;gSeelalso revised l;Changef01i06 K Which:provjdes;albroad; discussion off how;the; applicabilities foriCT523:1;1:1E3 lily 2l 3;1;3;51andl3?E 3;6;haveibeen revised.

l l- 02 01 'A In the conversion process this LC0 will be combined with the SDM LC0 applicable for T., > 200 F. in accordance with CPSES Description of Changes - CTS 3N.] 2a 7/29/98

)

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-4 APPLICABILITY: CP,DC 3.1.2 Core Reactivity (Comanche Peak & Diablo Canyon)

DOC 05-06-A JFD 3.1-2 CTS SR 4.1.1.1.2 ITS SR 3.1.2.1 Comment: The note to the core reactivity SR in the STS states that ".. predicted reactivity values may be adjusted (normalized) ...", while the note in the ITS states, ".. predicted reactivity values shall be adjusted (normalized) ...". The ITS use of the word "shall" is based upon the CTS use of the word. The Bases supporting this SR adds a parenthetical phrase stating "... normalization (adjustment, only if necessary)...", indicating that the STS wording is preferable. Using the word "shall" implies that an adjustment must always be may, regardless of the necessity. Adopt the STS wording to the SR 3.1.2.1 Note.

FLOG Response:

Consistent with the STS wording, the ITS and CTS have been revised to use the word "may" instead of "shall". DOC 05-06-A is used as the justification for the CTS change.

ATTACHED PAGES:

Enci 2 3/4.1-2 Enci 3A 6a Enci 3B 3 Enci SA 3.1-3 Enci 58 B 3.1-13 Encl 6A 1 Encl 6B 1 I

)

l

REACTIVITY CONTROLS SURVEILLANCE REQUIREMENTS (Continued)

c. When in MODE 2 with K,,, less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving i on.o9-A

reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each ,

fuel loading, by consideration of the factors of Specification i 05"A -

4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

e. When in MODE 2;with%,'S10, 3 or 4, at least onca per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by 0 0'*M '

consideration of the felic;;ing factors:

1) Rcactor Cociant Systcm bcron concentration  ; otatoc ;

2) Control rod position,

3) Reactor Occiant System overagc temperature.

4) Fuel burnup bescd on gross thccmal sacrgy gcncret4ent

5) Xenon concentration and

5) Sc;;rium concentration.

]

4.1.1.1.2 For Modes 1 and 2 only, the overall core reactivity balance shall be f05 + M ~ i compared to predicted values to demonstrate agreement within i 1% a k/k once prior tol. entering Mode 1 after.each r.efueling;andlat least once per 31 Effective Full E 054LS i ,

Power Days (EFPD)!thereafter This comparison shall consider at least those factors s'" " in Specification 4.1.1.1.1a. abovc. Tthe predicted reacti 05 06 A values sheHeay e aajusted (normalized) to enrrm nand to the act 0 3.1 4 conditio sladjustment;(normalization shall;;;- be'per. formed; prior to exceeding a fuel burnup of 60 EFPD after eac loading.y.If reactivity 5054LG- '

balanceiisinot within111mitst within:7< days 72_houE:;,.Levaluate the Safety Analyses. and1establi sh ; appropriate ; operating Restrictions; and/or ; surveillance 05 05 Ls requirements,lor.be;ini.at11 east,Modefwithinthe1nex.t!6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; TR 3.1 003 l

l l

CPSES Mark.up of CTS - 3N.1 3M 1-2 7/29/98 i

CHANGE NUMBER RSliC DESCRIPTION 05 06

~

m Net opplicabic et C"SES. 00: Cerc.cr;;icrc Cm. pori:;;r, A

T;ble (Er.cic are 2).CTSTSR A;1';1;5.1Trcylires:.that l7 I mummmmmmmun the: predicted; reactivity; values "shallibe;adjus.ted j

,( normaljzed) lag 60'EFPDiafterirefueling. TITS;SR13,;1(2.'1 states the normalization;astmay cbeladjusteds;This recognizes 1that; normal ization;i s;notEnecessaryij f predictedandlesasured:corelreactivitylareiwithin tolerancegJhelschedulinglofathej normalizationrof predicted:and; measured;coreireactivityicontinuesitolbe required at1601EF, PDC;Thereforenthis: change; reflects clarjficati_oq1of existing;1ntentland;isiconsidered adminj_strativei 0507 LS?24 Notiappl icablelat[CPSESis See;Conversionicompari son 7g 3,3,gg3 Table (Enclosure /38);

06-01 R The CTS 3.1.2.1. Shutdown Boration Flow Path and associated SR 4.1.2.1 are relocated to a licensee contro11edQThis is consistent with NUREG 1431. [Jhe 0 3.1 6 boration; subsystemc of;the chemical;andsvolumel; control system;(CVCS);provides;.thelseansitoimeet;onelof;the

_functiona1 Requirements (offtheCVCSH1;es,J.tocontro]ithe chemicaEneutronlabsorberl(boron) concentration;jnithe;RCS andito helpicontro12the;boroniconcentration:to _ maintain shutdown; margin:(SDM) GTo: accompl i sh' thi sifunctional requirementgthe;.borationisystems;TSirequirela; source:.of borated; water @ 7e:ottmore:flowjpaths;tolinjectithis borated; water %Jo;thereactorJcoolantisysteel(RCS) Rand appropriate; charging.pumpsitolprovjde;the snecessary charging; head; l ThisfproposedISjrevision;relocatesfrequirementstwhich;do notlmeetithe JSicriterial.in 10CFR50.36(c)(2)(11)Rto documentsiwith:e_stablished; control programsE This regulation addresses 1the: scope;and; purpose loffTSi tin doing(so21tIsetsiforthialspecjficjsetioflobjective criteria lforLdetermining whichtregulatory requirementsfand i operatingErestrjetions;should be included;in;the3TS) gelocation;of;theseirequirements; allows;the;TSito1be L reserved;only1forithose;conditionslorfl imitations tupon l reactor;opetation which areinecessarylto:obv1ateithe 1 j possibiljtyfof!an; abnormal [ situation'orfevent;giving rise

!' totanliemediateithreat:to:theTpublic; health;andJsafety therebydfocusjng theIscopeLofithe;TSHAnlevaluatio3:of?the appl icability; otthese [criterial to ;thi sispeci fication;j s provided;1niAttachment 21; Tofensureian, appropriate]leveltoficontrol;;;these requirementsMillibe;relocatedto;1).; documents;thatfare CPSES Description of Changes - CTS 3N.I 6a 7/29/98

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1 l Core Reactivity >

.3.1-9a l 3-1-3 3.1:2 l i

1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 - - - - --

NOTE - - ---

3.1;2;1 The predicted reartivity values may sheH may be adjusted (normalized) to correspo u tc um [.4 I i measured core reactivity prior to exceeding a l fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within Once prior to i 1% Ak/k of predicted values. entering H0DE 1 after each  !

refueling AND

..... NOTE --

Only required after 60 EFPD 31 EFPD thereafter 1

l I

CPSESMark-up ofNUREG-1431 - ITS3.1 3.1-3 7/2988

Core _ Reactivity SBN-( > 200 F B 3.1.2 a.3.1 5 BASES SURVEILLANCE SR 3-1-3 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design I calculations at B0C. The SR is modified by,a Note. The Note indicates requires %t the normalization;4;;; ;r,t,;nly i' l

r.;;;;;;rygofpredictedcorereactivitytotnemeasuieuvalue must oaKe place within the first 60 af M tive '"11 mwar days (EFPD) after puh fuel loadin HoweverRif5 the_deviatio 0 3.1 4 A tween; measured and predicted values is withintthe

( associated; measurement.andlanalyticalluncertainies 11tJJs

%essary;to1 normalize:the' predicted' core reactivity?Els allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1 is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (0PTR, AFD, etc.) for prompt indication of an anomaly.

REFERENCES 1. 10 CFR 50, Appendix A GDC 26. GDC 28, and GOC 29.

2. FSAR, Chapter [153 l l

i l

CPSES Mark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-13 S/ ISM 7 l

[ l

i JUSTIFICATION FOR DIFFERENCES FROM NUREG 1431 l l NUREG 1431 Section 3.1 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431 Revision 1, to make them plant specific or to incorporate generic changes resulting from the Industry /NRC generic change process. .

The change numbers are referenced directly from the NUREG 1431 mark ups. For l Enclosures 3A 3B, 4, 6A and 6B, text in brackets "[ ]" indicates the information is l plant specific and is not common to all the Joint Licensing Subcommittee (JLS)

plants. Empty brackets indicate that other JLS plants may have plant specific information in that location.

[ CHANGE NUMBER JUSTIFICATION

{

i 3.1 1 In accordance with TSTF 9 Rev 1, this change would relocate the t specified limit for Shutdown Margin (SDM) from the ISTS to the l COLR. This change occurs in several specifications including the

!~ Specifications for SDH and those specifications with ACTIONS that require verifying SDM within limits. j 3.1-2 The notc for ITS Sil 3.1.2.1 indicatcs that predicted f acctivity valucs may bc adjusted (ncrmalized) to correspond to the acasured cerc reactivity prior to cxeceding a fuci a.3.1 4 I

burnup of 60 !Ti'O after cach refueling. "cwcVer, both the Bases for ITS 3.1.2 and the current TS requircacnts in l Specificatica 3.1.1.5 state that the normalization shall be done prior to cxcccding a fuel burnup of 50 EFf'O after cach refueling.

Thereforc. the actc has bcca revised to indicate that this is a requiremcat. Notlused;  ;

)

! 3.1 3 Not Applicable to CPSES. See conversion comparison table (enclosure 68).

1 3.1 4 SR 3.1.4.2 of NUREG-1431. Revision 1 would be deleted. In

! accordance with TSTF 13 Rev 1, the intent of this SR is only to l determine the next frequency for SR 3.1.4.3. Performance of SR i 3.1.4.2 is not necessary to assure that the LC0 is met; SR l 3.1.4.3 fulfills that purpose. Therefore, SR 3.1.4.2 may be deleted. In addition, the Note in the Frequency column of SR 3.1.4.2 would be moved to become Note 1 in the Surveillance column of SR 3.1.4.3. This is for clarification purposes. As discussed in CN 3.19, section re numbering results in SR 3.1.4.3 of NUREG-1431, Revision 1 becoming SR 3.1.3.2.

l 3.1 5 Per current TS [3.1.3.1], the words "with all" have been removed from ITS LC0 3.1.4. This is a clarification that ensures the proper interpretation of the LCO. The change makes it clear that only one channel of DRPI is necessary to meet the alignment accuracy requirement of the LCO. With the word "all" in the l

CPSES Differencesfrom NUREG-1431 -ITS 3.1 1 7/29/98

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_ !ll'

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: 03.1-5 APPLICABILITY: CP REQUEST: 3.1.2 Core Reactivity (Comanche Peak)

ITS Bases 3.1.2 Comment: The ITS Bases pages have the incorrect title in the page headers (SDM vs Core Reactivity). Correct the headers to ITS Bases 3.1.2.

FLOG Response: The header for the markup of the ITS Bases 3.1.2 is corrected to be " Core Reactivity".

ATTACHED PAGES:

Enci5B B 3.1-9 thru B 3.1-13 l

l

4 Core Reactivity SOM d T,,, a 200#F B 3.1.2 l o.3,1 5 BASES -

BACKGROUND (continued) In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and in the fuel remaining from the previous cycle provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.

When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP.

Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.

APPLICABLE The acceptance criteria for core reactivity are that the SAFETY ANALYSES reactivity balance limit ensures plant operation is maintained within the assumptions of the safety analyses.

Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents ,

or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis

, evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical

! benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.

(continued)

CPSES Mark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-9 S/IS/97

Core. Reactivity S9M (, m 200T B 3.1.2 o.3,1 5 BASES APPLICABLE Design calculations and safety analyses are performed for each SAFETY ANALYSES fuel cycle for the purpose of predetermining reactivity behavior (continued) and the RCS boron concentration requirements for reactivity control during fuel depletion.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models

, used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (B0C) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate.

If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond B0C, or that an unexpected change in core conditions has occurred.

The normalization of predicted RCS boron concentration to the measured value i:; typieeHy shall_ be' performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle.

Core reactivity satisfies Criterion 2 of the NP,C Policy Stat =nt 10CFR50.36(c)(2)(11).

LC0 Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled altered once the l

core design is fixed. During operation, therefore, the LCO can l

only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of i 1% ak/k has

~

I (continued)

CPSESMark up ofNUREG-1431 Bases - ITS3.1 B 3.1-10 S/1S/97 l

j

Core: Reactivity SOM % s 200 i B 3.1.2 o.3.1 5 BASES LC0 been established based on engineering judgment. A 1% deviation (continued) in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.

When measured core reactivity is within 1% ak/k of the predicted value at steady state thermal conditions, the core is considered to be op2 rating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron i worth) before the limit is reached. These values are well within j the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.

APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERHAL POWER. .As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in H0 DES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing.

In H0DE 6. fuel loading results in a continually changing core reactivity. Boron concentration requirements (LC0 3.9.1, " Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).

I i

ACTIONS A.1 and A.2 l Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured 1

(continued)

CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-11 S/1S/97

ore;Reactivi.ty - 4 ,s 200 F B 3.1.2 0 3.1 5 BASES ACTIONS A.1 and A.2 (continued)  !

core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are I adequate for representation of the core conditions. The 4 required Completion Time of 7" days 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is bated on the TR 3.1 003 low probability of a DBA occurring during this period, and j allows sufficient time to assess the physical condition of l the reactor and complete the evaluation of the core design and safety analysis.

Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of ,

the RCS boron concentration requirements may be performed to l demonstrate that core reactivity is behaving as expected. If an '

unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to I provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power cperation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.

TR 3.1 003 The required Completion Time of 7 days 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is adequate l for preparing whatever operating restrictions or Surveillance j that may be required to allow continued reactor operation. '

kl If the core reactivity cannot be restored to within the 1% ak/k limit, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 LCO 3.1;1 Required ActioniA~11would occur. The allowed Completion Time is f reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

CPSESMark-up ofNUREG-1431 Bases - 1153.1 B 3.1-12 5/1S/97

6AA C E" Core: Reactivity mg ,'

BASES SURVEILLANCE SR 3-4-s r3:1'.2.1 REQUIREMENTS Core reactivity is verified by Priodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including contrcl rod position, moderator temperature, fuel temperature,' fuel' depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (adju.stmentilonly;jfinacessaryl of predicted core reactivity to the measured value mut take place within the first 60 effective full power days (EFPD) after each fuel loading. .This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1.

l 15 acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.)

for prompt indication of an anomaly.

REFERENCES 1. 10 CFR 50 Appendix A, GDC 26 GDC 28 and GDC 29.

2. FSAR, Chapter f153

{

CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-13 5/15M7

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-6 APPLICABILITY: CP, DC REQUEST: CTS 3/4.1.2 Boration Systems (Comanche Peak and Diablo Canyon)

DOC 06-01-R Comment: The Discussion of Change (DOC) needs to specify where the CTS specification is being relocated. Correct the DOC.

FLOG Response: DOC 06-01-R is revised and Technical Specification Screening Form for CTS 3.1.2.1 prepared to provide additionaljustification for the relocation. This justification shows that the boration system is not assumed to operate to mitigate any accifent. The maintenance of SDM provides all required reactivity margin. Since the system does not mitigate an accident, there is no installed instrumentation which is used to detect cr indicate a significant degradatiori of the RCS boundary. This system is also not associated with any variable, design feature, or operating limit that is the initial conditions of an event which challenges a fission product boundary. The boration system is not a part of, nor does it support a system requiring that support, to function as part of the success path to mitigate a design bases accident.

The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998.

ATTACHED PAGES:

Enci 3A 6a & 6b Att 21 1&2 i

1 1

CHANGE '

NUMBER N2fC DESCRIPTION t

05 06 A Not applicabic at CPSES. Sec Coraersica Cc;;;parisca 0 3.1 4 T;ble ' Enclosure 30). CTS'SRL4;1,135~.11 requires:that the, predicted ; reactivity; values?shall$ beladjusted l (normal i zed)lat ' 60 ' EFPD :after;; refuel i ng CiITS:SR L3 ;1' 2;1 . I states;the normaliz3tionfas;"may":beladjusted.1This I recognizes;that'normaliz? tion;isinot;necessary_1f i pr.edi ctedland; measured; core ; reacti vity:are Lwj thin J tolerancegThefschedulingloff the; normalization of predicted:and1 measured; core reactivity: continues;;to be requi red ;at:60 ; EFPDE Tnerefore,ithi sichange/ refl ects I clarificationiofdexisting;l intent andsisfconsidered administrative? l 05;07 LSt24 Nottapplicab]eiat;CPSES;iSee Conversjon; Comparison m-3.1-003 Table (Enclosure 38)]

06 01 R The CTS 3.1.2.1, Shutdown Boration Flow Path and associated SR 4.1.2.1 are relocated to a licensee -

contro11edAThis is consistent with NUREG 143 <The ~

03a4 1on; subsystem;.of the chemicaLand vol_ume,codrol system 2(CVCS);providesithe;means;tofmeetioneiofthe functionalfrequirementsioftheiCVCS'ilile;,toicontrolEthe chemical neutron; absorber 1(boron)ll concentration in;the RCS ,

andf to helpicontrolf thef boron 2 concentration to; maintain l shutdownJmargin;(SOM) U Tolaccomp1.ishithis1 functional.

requirementnthe;boraticn systems 1TSir.equireLa; source:of boratedLwater';onelorlmore flow? paths;to 1.njectthis borated: water;1_nto the; reactor; coolant system;(RCS) tand appropriate charging; pumps tofl provide _the necessary charging; head {

ThisfproposedjTSirevjsion'relocatesLrequirements',1whichdo notLeeet the;TS;criter,1alin110CFR50136(c)(2)(ii)R to  !

documents;withEestablished control programKThis  ;

regulation; addresses)thefscopeidnd purpose 20f LTSGIn doinglsoff;1tisets;forth~acspecific. set ofiobjective .

criteria;forl determining which: regulatory;regi11rements-and operating restrictions should be:jncludedli.nLthe  : TS:

Rel.ocationlof;theseirequirementsl allows the;TSitolbe reserved 1only;forithose;conditionsforjlimitations-upon reactot operation which are:necessary,to~ obviate:the I possibili.tyfof;anllabnormalJsituationiorieventlgiving rise to;an[jemedi. ate: threat;to thelpublic; health;and safety thereby focusing the: scope;;of;thefTSCAn" evaluation 'of the . ,

appl.icability:of these criteria tolthisispecificationis provided(iniAttachment;21]

ToLensure;an; appropriate 1evel;;oficontrol",Lthese requi rements will ibe; rel ocated ;to ri): docume.nt.s ; that . are CPSES Description of Changes - CT5 6a

}

CHANGE NUMBER HMiG DESCRIPTION subject;to:theprovisjons;oC101CFR;50:59Td2);other 0-3*1 6 licensee; documents;whichhave;similarfregulatory control sE(e. g.%the; Qual ity;AssurancelP,lan gas describedlin;the;FSARCwhich!1sicontrolled;by 10CFR50;54a)Aor3); to; prograssj.that;arelcontroll ed ; via thel: Adej nistrative_ Control s:sectionl ofithe;1mproved tTS;I Thei1.denti fjcation;ofathelspeci ficil icensee ; controlled document containing;this;requireeentjjsiprovidedfin Enclosure;38.loGthe ' conversion; submittalg Compliance l wjthithelrei ocatedirequj rementsiw1111 notLhe affacted byXthjs M oposed; change;to;thelcurtent: Technical Speci fications.;IThelrequired: periodicisurvejil ances;will continuelto be:performedit( JAsurelthatl11mitsLon patasetersjareimaintained M Thereforen relocationjofLthese requirements,willihayeino;1mpaction.! system operability;oc the: maintenance:of;;controll ed: parameters;with ,

07 01 R The C , Oper ation Flow Path and associated SR 4.1.2.2 are relocated to a licensee i controlled document. This is consistent with NUREG 1431.

The;borationisubsystenfof the chemicalEand;voluse 0 3.1 7 contro11 system'(CVCS)1provides;the:meansjtolmeetsone ofdthe;functjona11 requirements;ofithelCVCSE ije m to controlithelchenjcalineutron;ab.sorberi(boron)

)

concentration:jn~the;.RCS"and;to; help;contro11the boron l'

concentrationito saintainishutdown;margini(SDM).M s To accomplj sh ;thi s: functional irequi rement Cthe; boration systems]TS;requitefalsource;ofborated;WaterRone;ormore i

flow;pathsitolinjectithis! borated; water;jntolthe; reactor coolantisystem;(RCS)Randlappropr.iate; charging: pumps;to Provide;the;necessaryicharging; head; Thi s: proposed:TSitevj sior.1 relocates ; requirements which;do not'l meet;.thelTSicciterjalin 10CFR50J36(c)(2)(11)~,3to documentsfwithlestabl i shedj contro11 prograns gThi s

tegulation; addresses 1thelscopelandipurpose

of;TS C In

! doingisoZitJsetsiforthia:specjficisetloflobjective criterja;forl determining which regulatoryirequirementsiand operating 1 restrictions [shoul.d; belincl uded ;inithe;.TS ?

Relocation;ofdthesepequirementsiallowsn theJS;toLbe resetved:only;for;those;conditjonslotiljoitationslupon reactor; operation;whichjere1necessary;tolobviateithe possibility;oflan; abnormal; situation;or eyent:giyinglrjse to;an]immediate; threat:toithe;public1healthiandis_afety thereby;focusingjthe; scope;ofithelTSTAnl evaluation'of the appl icabil ityiofi theseictiteti a;tolthj s: speci fication ; i s provided;in: Attachments 21; l

CPSESDescription ofChanges - CTS 3N.1 6b 7/29/98 f

L-

I

' l I

TECHNICAL SPECIFICATION SCREENING FORM l

l (1) TECHNICAL SPECIFICATION 3.1.2.1 BORATION FLOW PATHS - SHUTDOWN 1

Applicable MODES: MODES 5 and 6 I

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

l

_ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or

, transient that either assumes the failure of or presents a challenge to l the integrity of a fission product barrier.

_ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has slown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification (TS) shall be retained in the TS.

If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document.

(3) DISCUSSION The Bases for this limiting condition for operation (LCO) state that the purpose is to assure negative reactivity control is available during each mode of facility operation.

The boration subsystem of the chemical and volume control system (CVCS) provides the means to meet one of the functional requirements of the CVCS. i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain shutdown margin (SDM). To accornplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the reactor coolant systern (RCS), and appropriate charging pumps to provide the necessary charging head.

The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron citution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the tx> ration subsystem is no' assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel Attachment 21 l

d sign limits will not be exczeded for normal shutdown end anticipated op: rational occurrenc:s. Tha SDM d:fints tha d: gree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

l Based on the foregoing, the boration subsystem is not installed instrumentation that is l used to detect or indicate a significant degradation of the reactor coolant pressure

! boundary (RCPB); therefore, this TS does not satisfy criterion 1.

i The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or l challenges the integrity of a fission product barrier. Therefore, the boration subsystem l TS does not satisfy criterlun 2.

For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3.

The boration systent function at shutdown is used to provide reactivity control, in particular for dilution events. As part of the CPSES Safety Monitor shutdown model l development dilution events as an initiator to boiling or core damage were investigated.

This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other plant shutdown models, including the Outage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is concluded to be an insignificant contributor to shutdown risk for CPSES.

Given the foregoing, it is concluded that the boration flow path SSCs oro shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy criterion 4.

(4) CONCLUSION

_ This Technical Specification is retained.

.X. The Technical Specification may be relocated to a licensee controlled document.

l I

> 1 l

]

1 3

Attaclunent 21 l 2

)

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-7 APPLICABILITY: CP, DC i i

REQUEST: CTS 3.1.2.2 Flow Parth - Operating (Comanche Peak & Diablo Canyon)

DOC 07-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DOC. A relocated screening form is not provided for this relocated specification.

FLOG Response: For CPSES, DOC 07-01-R has also been revised to provide additional justification for the relocation.

The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telept'one calls on June 25 and June 30,1998.

For Diablo Canyon, License Amendment (LA) 120/118 (dated 2/3/98) eliminates this RAl. LA 120/118 relocated ten TSs in accordance with 10 CFR 50.36. Thus, CTS 3.1.2.2 no longer exists and DOC 07-01-R is not applicable to DCPP. CTS 3.1.2.2 was relocated to ECG 8.4, which is part of an NRC approved program controlled by 10 CFR 50.59, and discussed in FSAR Chapter 16.

ATTACHED PAGES:

Enct 3A 6b & 6c Att 21 3&4 i

l l

CHANGE NUMBER EC DESCRIPTION subjectito;the2 provi sior,siof;101CFR;50 iS922); other 03'16 11censeeldocuments Lwhichi hayel,ssist]aKregulatory controls 3eigRtheiQualitylAssuranceiPlan;Eas desetibed;in;the1FSAREWhich7jscontro11ediby 10CFR50.54a)Eori3);to; programs;thatIare:pontro))edivia thel Administrative; Control s; sectionTof;thejjaprovediTS 3 TheltdentifjcationlofLthetspecificilicensee: controlled document:contatning;this; requirement 11s;provided:1n Enclosure;38:ofsthe;conversj on ~ submitta]l.3 Compliance 1with:the:rel.ocated;requirementsiw11Enot be affacted,byjthjsl proposed;changelto;the:currenttTechnical Specification.sEJhe;.requiredj perjodic;survejll ancesiwill continueltolbe; performed;to: ens _ure;that;11mitslon )

Parameters;areimaintained E Therefore.Trelocation of;these requirements;wjlllhave;no; impact;on;systenloperabilitylot thelmaintenanceioficontrolledfparameters Within;11mits;  ;

07 01 R The CTS 3.1.2.2 Operating Boration Flow Path and associated SR 4.1.2.2 are relocated to a licensee controlled document. This is consistent with NUREG 1431. A boration; subsystem:of the chemical and;v51ume coni; roe system 1(CVCS): provides jhej usahs;toimeet;one ofdhe; functional! requirements 1ofithetCVCS;?t;p M to l l

controlltheic.hemicalzneutron;ab_sorber;(boron) concentrationiin ;theLRCSland ito lhel p;contro12the; boron concentration 1to maintain [shutdownimargin1(SDM) M To accomplishf this1 functional!requirementEthe;.boration systensiTS; require?aisource;ofboratedl Water;Tonefor;more l flow; paths;toljnjectithisiboratedjwaterijnto the: reactor cool ant lsystemi(RCS) C and ; appropriate; charging; pumps ;to l provide:the necessary; charging < head!

ThisproposediTStreyisionirelocatestrequirements7;whjch,do notineet'; the ,TS;; criteria:lin ;10CFR50 ;36(c) (2) (ii)nto documents; with;establ ished;controlf programs nThi s

! regulation; addresses;thelscopeland purposeloflTSGIn doingiso Rj t { sets;forthia (specj fic : set iofcobjective crjteriafforldeterminjng;which; regulatory; requirements and operating 1 restrictions;shoul d ibelincl udedfj n ;the;TS ?

Relocationiofithese; requirements; allows;the1TSLto be reservedLonlydorithosejconditionsZorjjimitationsiupon reactotloperagon1which7are.necessary;to_obvjate;the possibilityiofian:abnormalisituationEot:eventigiving rise toianlimmediatelthreatitolthelpublicjhealth and1 safety therebylfocusjngithel. scope of;the;TSfiAnjevaluation oftthe applicability lof;these criteriaitoithis . specification ;is provided;in!

Attachment:

217 CPSES Description of Changes - CTS 3N.1 6b 7/2988

CHANGE NUMBER RSE DESCRIPTION ToJensurelaniapproprjate:1eveRoffcontro1Rthese requirements;willibefrelocated;toJ1); documents;that:are

'ubject;tolthe prov131onsIof610;CFRl5.0j59,X2);othet s

l03.17

] icenseeldocuments;which;havelsimil atj regul atory controls 1(e;g7,;the QualityiAssuranceflan;iasidescribed

, in1the fSAR,1whichlis;contro11 edlbyJ0CFR50.; 54a) E or 3);to programsithatiarecontro]1edlyialthe; Administrative Controls ~section:ofitheLimptoved5TS C The11 identification oCthelspecifichlicen.see; controlled documentfcontaining thi slyequj rementij s] provided11nj Enclo.sure;3Blofithe conversionjsubmittalf.1 l

Compl jance ;with;the: relocated : requirements ;w1] Enotibe affectediby_this!ptoposedichangelto;theicurrentiTechnjcal Speci fications gThe;ryquired: perj od.iclsurveill ances;will, continuetolbe? performed:toiensureithatlljaitsLon parameters!are: maintained G ThereforeC relocation:of;the.se requi rements; w11E ha.vef noj impact; opisystem ;operab1] ity; or, the1 maintenance:of' controlled: parameters 1within;11mits;

< 07 02 A m acce danw with NUREk431, the limitation on charging pumps in Mode 4 has be moved to improved TS LC0 3.4.12.

08 01 R The CTS 3.1.2.3, Shutdown Charging Pump and associated SRs l 4.1.2.3.1 and 4.1.2.3.2 are relocated to a licensee

! controlled document. This is consistent with NUREG-1431. The: charging; pumps:are; components;withinithe 0 3.1-8

Boration System &Thejfunction

ofithelBoratjon Systes 1sito:ensureithat;negativepeactivity;controllis

! availab]eiduringleachinodeiotfacility; operation;ri l This propo_sediTSirevjsi_on; relocates [ requirements',Whjch:do

! notiseet;the:TS;criterjalin;10CFR50;;36(c)(2)(ii)nto

documents;withf.establ i shedicontrol; prograns NThis l regul ationf; addresses tthe

. scope;'and; purposelof;TSliIn doinglsoQtsetsfforth:a specifjcisetiof; objective criteria;for; determining which; regulatory l requirements:and oper.atingtrestrictionsishouldibeljncludeddi n;theJST Relocation:ofitheseirequj rements3110wsithe1TSitoibe reserved lonlyzfoGthose: conditj ons]!ril imitationsi upon reactorioperation;whichlare necessatyZto;obviateithe po.ssibil ity;ofi an: abnormal; si tuation Wdevent: giving;ri se to; antimmediate; threat' toj thelpubl icj heal th?c% safety thereby;focusj ng;theiscope;;of;theJS EAn; eval uationiofithe app 11 cab 111tylofithese; criteria;to this; specification;js provjdedlin: Attachment 217 Tolensurelaniappropriatellevellof;contro1Ethese requirementsiwjlllbe relocated;to'1)Jdocuments;that are subject to;the;prov.isionsfof,10iCFRT50 59E 2) other CPSES Description of Changes - CTS 3N.1 6c 7n988

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.2 BORATION FLOW PATHS - OPERATING Applicable MODES: MODES 1,2,3 and 4*

(*A maximum of two charging pumps shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F except when Specification 3.4.8.3 is not applicable.)

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification (TS) shall be retained in the TS.

If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document.

(3) DISCUSSION The Bases for this limiting condition for operation (LCO) state that the purpose is to assure negative reactivity control is available during each mode of facility operation.

The boration subsystem of the chemical and volume control system (CVCS) provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain shutdown margin (SDM). To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the reactor coolant system (RCS), and appropriate charging pumps to provide the necessary charging head.

The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of 1

Attachment 21

! 3 l

t

ths boration subsyst:m is not assumrd to mitigata this ev:nt. Furth rmore, R:f. 3 notis that th3 normal capability to control r: activity with boron is not cr:dited in tho accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of subcriticality thet would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

Based on the foregoing, the boration subsystem is not installed instrumentation that is used to detect or indicate a significant degradation of the reactor coolant pressure boundary (RCPB); therefore, this TS does not satisfy criterion 1.

The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2.

For these events, the primary success path for mitigation includes isolating the dilution fiowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that eb.her assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterien 3.

The boration function of CVCS at operation is explienly modeled in the CPSES Individual Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE  !

model of this function includes both the boration path from the boric acid transfer tanks via the boric acid transfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given the low failure probability and the redundancy and diversity of these flow paths, based on the results of the CPSES IPE.

In addition, because the ECCS function of injection from the refueling water storage tank via the CCPs ( and therefore one of the emergency boration flow paths) remains part of the improved tech specs in another section, the emergency boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

2L The Technical Specification may be relocated to a licensee controlled document.

I l

Attachment 21 4

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-8 APPLICABILITY: CP, DC I REQUEST: CTS 3.1.2.3 Charging Pump - Shutdown (Comanche Peak & Diablo Canyon) l DOC 08-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DO C.

FLOG Response: DOC 08-01-R is revised and Technical Specification Screening Form for l

CTS 3.1.2.3 prepared to provide additional justification for the relocation. This justification shows that the charging pump system is not assumed to operate to mitigate any accident. The response for a boron dilution event would be to secure appropriate valves prior to loss of SDM.

Since the system does not mitigate an accident, there is no installed instrumentation which is used to detect or indicate a significant degradation of the RCS boundary. This system is also not associated with any variable, design feature, or operating limit that is the initial conditions of l an event which challenges a fission product boundary. The boration system is not a part of, nor does it support a system requiring that support to function as part of the success path to mitigate a design bases accident.

The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998.

l ATTACHED PAGES:

Enci 3A 6c & 6d Att 21 5&6 l

l i

CHANGE NUMBER HSiG DESCRIPTION l l

Tolensurelanlappropriate11evel(of;contro]2the.se requitenents;W1111beitelocated;to21);documentsithat are 0 3.1 7 subject;tolthelprovisjon.si.ofl101CFR;50. 59E2)[othet ljcensee3 documents;whichlhaveisimilariregulatory controlsne;ggthe1Qua]ity;As.surance Planzasidescribed i jnitheJSARRwhichMsicontrolled!by210CfR50.54a);[orf3)1to programs;thatare; controlled 1vialthe; Administrative Contro1sisectjon of;the;1mprwved!TSTMheiidentification otithelspecificilicensee1 controlled idocumenticontaining thi s; requirement;is : provided11n l Enclosure!381ofithe conversionisubmittalij i

i ComplianceMthithe; relocated:requirementsiwillinotibe  !

affected byithisipropos.ed:changeltolthe: current; Technical.

SpecificationspLThelrequiredi petiodicisufiejihLnces;will_ g l

continueltolbelperformed jolensure;thatilimitslon j pataneterslarelmaintained E Therefore n relocationLofithese requirements;w1))ihave;nosimpaction;systemioperabilityLor thelaajntenance::oficontrolledl parameters:withinMajtsl I

i l 07 02 A In accordance with NUREG-1431, the limitation on charging pumps in Mode 4 has be moved to improved TS LC0 3.4.12.

l 08 01 R The CTS 3.1.2.3, Shutdown Charging Pump and associated SRs 4.1.2.3.1 and 4.1.2.3.2 are relocated to a licensee

! controlled doenmant. This is consistent with NUREG-l 143 % charging:pumpslare; W [withinTtne ation Systes E The; function [otthe Boration; System W

i js;tolensureithatnegative: reactivity;contro]Ejs aya11ableidurjng'eachinodelof; facility;operationflig i t

This;proposedjTS1 revision; relocates'requirementsEwhichido notimeetitheiTS crjterialinL10CFR50]36(c)(2)Oj)Rto documents; withiestablishedicontroli programstThi s regulation [addressesithelscopeland: purpose;of;TSigIn doing isoRitisetsiforthla 'speci ficisetiof;objecti ve etiterialfor; determining;whichzegulatoryirequirementsiand operatingitestrfctionsishould beZincluded;in;the;TSJ Rel ocationlofdtheseltequirementsja'1]owsj thelTS1to; be reserved;onlylfore those conditions 4ralinitations;upon reactorloperation1whichl are;necessary;tolobviateithe possibilitytofsanlabnormallsituation orfevent;giving rise l tolaniimmedjate;threatitoithe!publjcihealt.hiandjsafety thereby;focusjng;the; scope;of;the3S""Anievaluationtof(the applicability:ofitheseicriterialtolthis;specificationiis provideoEinAttachment 217 To:ensurelan; appropriate levellofocontrolf,1these requirements 1]11belrelocatedito[1)idocuments;thattare subjectito;the; provisions;ofJ1_0;CFR;50159,t2);othet CPSES Description of Changes - CTS 3N.1 6c 7/29/98

CHANGE NUMBER N2iG DESCRIPTION licerrsee: documents;Which; hayelsimil_ arc regul atory controls:(e:ggitheQuality: Assurance:PlanEas 0u8 described;1n;theJSAREwhichlis contro11ediby 10CFR50J.54a):, lori 3)'to programs;that:arel controlled _ via theIAdmini strativeiControl s :section :ofithelimprovediTS ;]

Theitdentificationiof;thelspeci fic.j J centee; controlled documentIcontaining thi_s; requirement:2jsiprovidedli.n Enclosure 238 of;the;conversioMsubmittal:3 Compljancelwithithe:relocatedirequirements:will not;be affectediby,,this pryposedichangelto;the; current; Technical SpecificationsEThe;requiredjperiodic;surveillancestwill continueltolbe performeditolen.surejthat;11mits2on Parameters!are: maintained C Therefore g relocationlof;these requi rements1 willi hayelno depactiondsystaeloperabil ityiot the i

(~;maintenanceiofcontrolledjparametersjwithinilimits?

08 02 H Consistent with NUREG 1431, tne surveinanWieyuirements for charging pumps when below 350 F have been moved to SR 3.4.12.2 of the improved TS and the frequency has been increased from once per 31 days to once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

08-03 LS-19 Consistent with NUREG 1431, the metiiod of verifying that charging pumps are incripable of injecting into the RCS has been deleted.

09-01 R The CTS 3.1.2.4, Operating Charging Pump and 0 3.1 10 associated SRs 4.1.2.4.1 and 4.1.i. 4.2 are relocated to a licensee controlled document. This is consistent with NUREG 1431. Thelchargingfpumpsiare componentswjthin,the'Boration;SystenNThe'functionDf theBoration'. System;isito; ensure that1 negative; reactivity;controEisiayailableiduring;.each modelofifac111.tyioperation;3 This;proposedtTS revisionfeloca.tes; requirements,Swhich;do notimeetithelTSicrjterialin:10CFR50;36(c)(2)(ji)',3to documents;With lestabli shedlcontrol" progranss Thi s ,

regulatjon: addresses 1thelscope:and purpose ofiTSMIn

! doing3sosjtisets forthlaf:specifjciset;of' objective l

crj teria:foridetermining ,whichiregul atoryirequi rements; and operatingfestrjetionsishoul.dibe.; included in;the;TS; Relocationiof these i requirements lallowsitheJTS toLbe reseryed; onlylforj those1 conditions;oCl imitationsj upon reactorioperation whichiareinecessarycto obviate!the possibility lofian! abnormal;situationiorlevente91.ving rise tolan:1mmediatelthreatito:thelpublic1 health andisafety therebyfocu.sj ngithelscope tofitheHS7:An' evaluation ofithe CPSES Description of Changes - CTS 3N.1 6d 72988 l

L_____--_-_-__--__-_.-__

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 1.2.3 CHARGING PUMPS - SHUTDOWN Applicable MODES: MODES 5 and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ X (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ X (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS.

If the answer to all four of the above questions is "NO", the TS may be relocated to a control led document.

(3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. Equipment required to perform this function includes: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power source from operable diesel generators.

The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate

( valves in the reactor makeup system before the SDM is lost. Operation of the boration

subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

! Attachment 21 5

Th3 boration subsyst:m TS is not epplicabla to install:d instrum ntation us:d to d tsct or indicat3 a significant d gradation of ths RCPB; thirsfora, this TS does not satisfy criterion 1.

The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2.

For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a SSC that is part of the primary success path and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a finion product barrier; therefore, the TS does not satisfy criterion 3.

The boration system function at shutdown is used to provide reactivity control, in particular for dilution events. As part of the CPSES Safety Monitor shutdown model development, the issue of dilution events as an initiator to boiling or core damage was investigated. This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other plant shutdown models, including the Outage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is concluded to be an insignificant contributor to shutdown risk for CPSES.

Given the foregoing, it is concluded that the boration flow path SSCs are shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy enterion 4.

(4) CONCLUSION

,_ This Technical Specification is retained.

.2L The Technical Specification may be relocated to a licensee controlled document.

I Attachment 21 6

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-10 APPLICABILITY: CP,DC REQUEST: CTS 3.1.2.4 Charging Pump - Operating (Comanche Peak & Diablo Canyon)

DOC 09-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DOC.

FLOG Response: DOC 09-01-R is revised and Technical Specification Screening Form for CTS 3.1.2.4 prepared to provide additional justification for the relocation. CTS 3.1.2.4 is provided to assure negative reactivity control during MODES 1,2,3, and 4 operation. This justification shows that the charging pump system is not assumed to operate to provide negative reactivity control to mitigate any accident. The response for a boron dilution event would be to secure appropriate valves in the reactor makeup system. Since the system does not mitigate an accident, there is no installed instrumentation which is used to detect or indicate a significant degradation of the RCS boundary. This system is also not associated with any variable, design feature, or operating limit that is the initial conditions of an event which challenges a fission product boundary. This subsystem in the boration system is not a part of, nor does it support a system requiring that support to function as part of the success path to mitigate a design bases accident.

The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998.

ATTACHED PAGES:

Enci 3A 6d & 6e Att 21 7&8 l

l i

CHANGE NUMBER HSiC DESCRIPTION licenseeldocumentsfwhich:haye:similarf; regulatory 0 3.1-8 controJs1(e;g a the Quality:Assuranceflantas

{

describedjin'the;FSARi!.which!1sicontrollediby j 10CFR50154.a) Eor;3)f to; prograssithatiare;controll ed ; v ia j the; Administrative;Controisisection;off thetimproved:TSU The !;j denti ficationlofdtheispeci f J c;11censeeicontrolled document;conta.ining:thistrequirementtisiprovidedlin ,

Enclosure;;38 of the:conyersion1 submittal'.2 Comp 11anceiwithithe; relocatedirequi rements ;will;; not; be affectedj by;thi s;;proposedichange, to;the] current lechnical.

SpecificationsMThe;requi red; peCiodic:surveill ances;will continuelto;bejperformed;to:ensureithattlimitslon parameters l areimai ntainedilTherefore grel ocationiotthese requirements;willihavelno; impact;onisystem; operability;or the; maintenance;of[ controlled: parameters within111mitsJ 08 02 H Consistent with NUREG 1431, the surveillance requirements for charging pumps when below 350*F have been moved to SR 3.4.12.2 of the improved TS and the frequency has been increased from once per 31 days to once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

08 03 LS 19 Consistent with NUREG 1431, the method of verifying that charging pumps are incapable of injecting into the RCS has been deleted.

09 01 R The CTS 3.1.2.4, Operating Charging Pump and b 0 3.1 10 associated SRs 4.1.2.4.1 and 4.1.2.4.2 are relocat to a licensee controlled docume t Thh j l}

consistent with NUREG 1 - .he charging. pumps;are

BorationLSystes

dlThejfunction;of

- ;-

;; Z R ,_

theiBoratj on lSystesfj s, toj ensure thatinegative:reactivjtylcontro1Ms:available during;each modelofdfacilityloperationu This;proposedLTSireyisionrelocat_esjrequirementsawhichLdo

l. not;meetithe!TSicrjterialin 10CFR50:36(c)(2)(11) ito documents; withlestabl i shed: control; prograssEThi s regul ationiaddresses t,the; scope;andJ purposelof[TS ? ; In doing ;so,S itlsetsjforth ;alspecifj cise_t; ofjobjective crjter1alfor;determiningWhich;regulatoryirequirements;and operating restriction _sishoul.d belincludedrintthe!TSI c Rel ocation;of2these Tequi rements;all ows[thelTSi to;be

( reservedionly] forithoseicondi tions;or;l imitations L upon reactorloperationiwhichlare:necessary,toLobviatelthe possibility;ofianjabnormalJsituationforl event;giving rise to;an:immediatelthreatito;thelpublic;-healthand; safety ther.eby; focusing;thelscopeLotthe3STEAn; eval.uationoflthe CPSES Description of Changes - CTS 3N.1 6d 7/29/98 ,

i L ____ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ - - - - - - .

CHANGE NUMBER NSE DESCRIPTION app 11cabifRy;of2these: criteria {to;this speci ficationli s i;provided11 n: Attachments 21] 0'3 l'10 l

Tolensure':an[appropriatejleyeliof"contro1E

these requirements;willlbe;relocateditol1)1 documents;thatiare subject;tolthe; provisions;ofL10;CFR;5015922);other licensee; documents 1whichlhayeisimilariregulatory; control _s (elg?;;the'Qualjty Assurancefl an t[as;descr,1 bed iinithe l

FSARawhichlisicontro))ediby; 10CFR50] 54a)E orf 3)3o program.sithatiare;contro)]ediviaithefAdminj strati.ve Contro] sisection soff the%1mproved iTSLLThelidenti fication

]

ofithelspecific;]jcensee! control]ed;documenticontaining thisitequirementijs:provjded[in1 Enclosure 3B;off the

. conversion 1submittalg i

Compljance wjth:thelrelocated; requirements;wil);not;be affected;byithis; proposed l change:toitheicurrent1 Technical l Specif1cationsimThe: required:petiodic:sutveillances:will .

continueltolbe; performed;tolensure thats11mitslon parameters;are , maintained G3herefore trel_ocation;of1these requirements liw11Ehave nolimpactJontsystem;operabilitylor theina.intenance~oficontrolled, parameters;within;]1mits; 10 01 R The CTS 3.1.2.5, Shutdown Borated Water Source and associated SR 4.1.2.5 are relocated to a licensee controlled document. This is consistent with NUREG 1431.

The borated; water; source 11slaicomponentiwithinEthe 0 3.1 11 Boration.;SystesMThe] function:ofithe;Boration System]

to ensure Jhatlnegative feactiVity;controliis avail able; duringleachinodelofi facil i ty! operation .i Thi siproposed tTS; revi.sj onfrelocates Requirements gwhich [do not;meetithe;TS? criteria;in;10CFR50;36(c)(2)(ii)TCto documents 1withlestablishedicontrolsprogramsRThis tegu]ationladdressesitheiscopelandj purposeiof;TS EIn doing ? so C itisets;forthialspeci ficiset[ofiobjective l criterialfor determining;which;regul_atoryirequirementsfand m

operating:restrictionsishouldibelincludeddnitheTS.'

RelocationiofltheseirequirementsiallowsitheiTS1to;be reserved;only forithoseiconditj ons;orJ11mitationslupon l reactorLoperation;which!are~necessary;to; obviate;the possibility;ofianiabnormal isituationforl event; givingiti se to;anlimmediateithreat'tolthe:public;healthiandisafety thereby; focusing;the;scopeiof;the;TSEAnleva]uation ofithe appl icabil ity:ofitheselcriteri a itoithi s1speci fication;j s provided;in Attachments 211 Tolensure?aniapproprjatellevellof control:,%these requirements willibelrelocat_ed; toil); documents;thatiare CPSES Description of Changes - CTS 3N.1 6e 7/29/98

l l TECHNICAL SPECIFICATION SCREENIN3 FORM l I

(1) TECHNICAL SPECIFICATION 3.1.2.4 CHARGING PUMPS - OPERATING Applicable MODES: MODES 1,2,3* and 4* "

(*The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the charging pump declared inoperable pursuant to Specification 3.1.2.4 provided the charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375*F, whichever comes first.

"In MODE 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.)

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure ,

boundary. l

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety.

1 if the answer to any one of the above questions is "YES", then the TS shall be retained in the TS.

If tne answer to all four of the above questions is "NO", the TS may be relocated to a controlled document.

(3) DISCUSSION J

The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The equipment required to perform this function includes: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from operable diesel generators.

The boration subsystem of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron)

} concentration in the RCS and to help control the boron concentration to maintain SDM.

To accomplish this functional requirement, the boration systems TS require a source of I borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head.

Attachment 21 7

The boration subsyst:m is not assum:d to operate to mitigate the constqu:nc:s of a DBA  !

or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by me% oiedjustments to the RCS boron concentration.

Based on the foregoing, the boration subsystem is not installed instrumentation that is used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The boration subsystem TS is not associated with a process variable, design feature, or I operating restriction that is an initial condition of an event that assumes failure of or f challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2. ]

For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is j intended to regain the required SDM. This is desirable, but beyond the scope of a primary j success path action. The boration subsystem TS does not apply to a system that is part of 1 the primary success path, and which functions to mitigate a DBA or transient that either l' assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the refueling water storage tank (RWST) and associated flowpaths is required as part of the emergency core cooling system (ECCS) TS.

For the main steamline break (MSLB) event, the sequence of events takes the plant to cold shutdown conditions and; therefore, boration of the RCS is necessary. However, the boration flowpath in this case is required as part of the ECCS function.

The boration function of CVCS at operation is explicitly modeled in the CPSES Individual

' Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE model of this function includes both the boration path from the boric acid transfer tanks via the boric acid transfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given the low failure probability and the redundancy and diversity of these flow paths, based on the results of the CPSES IPE.

In addition, the ECCS function of the centrifugal charging pumps is explicitly modeled in the CPSES IPE and is being retained in the CPSES ECCS Tech Spec. Because the ECCS function of injection from the refueling water storage tank via the CCPs, and therefore one of the emergency boration flow paths, remains part of the improved tech spec, the emergency boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy criterion 4.

(4) CONCLUSION

_ This Technical Specification is retained.

1 The Technical Specification may be relocated to a licensee controlled document.

Attachment 21 8

t_____________ _ _ _ _ _ _ _ _

j

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL NFORMATION NO: Q3.1-11 APPLICABILITY: CP,DC REQUEST:

CTS 3.1.2.5 Borated Water Source - Shutdown (Comanche Peak & Diablo Canyon)

DOC 10-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DOC.

FLOG R.esponse: DOC 10-01-R is revised and Technical Specification Screening Form for CTS 3.1.2.5 prepared to provide additionaljustification for the relocation. This justification shows that the boric acid storage system is not assumed to operate to provide negative reactivity control to mitigate any accident. The response for a boron dilution event would be to secure appropriate valves in the reactor makeup system. Since the system does not mitigate an accident, there is no installed instrumentation which is used to detect or indicate a significant degradation of the RCS boundary. This system is also not associated with any variable, design feature, or operating limit that is the initial conditions of an event which challenges a fission product boundary. This subsystem in the boration system is not a part of, nor does it support a system requiring that support to function as part of the success path to mitigate a design bases accident.

The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998.

ATTACHED PAGES:

Encl 3A 6e & 6f )

I Att 21 9 & 10 l

I

CHANGE HUMBE8 RSBC DESCRIPTION pppl icabilityf ofithese ' criteria ;toithi s specification;j sj provided;in!

Attachment:

217 0-3.1 10 To; ensure;an appropriate 11evelfoficontro10these requirements;Wil]Ibe;relocatedjo:1): documents 1that are subjectLtoithe; provj sionsf ofl10!CFRl50. 59;E2)1other

]icenseeldocumentstwhichihaveisimilarJregulatory; controls (e;galthe Qualjty;AssurancelPlan;Easidesctibedfjnithe FSAR EWhich ;isicontroll ed ;byr.10CFR50; 54a)Eprf3);to Programs;that are; controlled;viatthe! Administrative Control s;section;ofitheitsprovediTS;'5Jheljdentification of;thelspeci fic Hicensee? control.ledidocumentT,containing this; requirement;1s^provided;in;

Enclosure:

38 ofEthe conversionisubmittalE Cpap11ance;withithe; relocated;; requirements)will not;be affectedihyithis;proposedichangelto;.the; current 1 Technical Specification.sELTheirequired; periodic lsuryeillances will contjnuelto ;.belperformeditolensurelthat11,1mitsf on parameters 1are ;maj ntained:RTherefore E relocationlof;these requi rementsLW1112 havelnolimpact' on ; system;operab1]1ty;ot theimaintenancelof[controll ed; parameters 3rithin tl imit s .?

10 01 R The CTS 3.1.2.5. Shutdown Borated Water Source and associated SR 4.1.2.5 are relocated to a licensee co ri document. This is con *+=t 'ith em-

~ ;botatedlwatercsource;1sia1componentLwithin the n 3,1,11 Boration1 System & The1 function pfithe Botation;Systemj toiensureithat;negativelreactiyityicontro1Jis availableiducingieachlsode;of;facilityloperationg This; proposed;TSireyision;relocatesErequirementsriwhich do potLeeet thelTSicriterialin110CF8tiO;36.(c)(2).(11)Rtp documents ,withlestabl i.shed ;controliprograns MThi s regulation; addresses;thelscopeiandLpurpose:of 2 TSEIn doing ;soaitisets;forth?aispeci ficj setiofiobjecti ve criterialfor"detemjning which1 regulatory; requirements and i operating yestrictionsishouldj belincl uded.11n ithelTS .:

1 RelocationlofEthese: requirement.s allows;the:TS;to be i resetyedionlyiforithose conditions;or[ limitations lupon j '

reactorioperationiwhich'areinecessaryltolobviateithe l possibilityofaniabnorma]isituation:orfeventJgivingl rise toianJjumediate2 threat 1toithelpublj cihealth ;and[ safety thereby1 focusing;the!. scop 3' of; the!TS;IAn; eval uationiofc the applicabil itylofj these;crj terj a;to; thi s: speci ficatj on yj s Proy1ded;in' Attachment;21?

To ensure laniappropriate11eyelfof controlMthese requitementsw11];belre)ocated~to:1)1 documents;that;ar CPSES Description of Changes - CTS 3M.1 6e 7a988 I: ,

t i

l 1

i CHANGE NUMBER ESE DESCRIPTION subjectitolthe; provisionsiof ~10lCFR;50. 5962)?other 0 3.1-11 licensee ldocumentsiwhi.chhave;similar; regulatory controls!(e;g,71thelQualityfAssurance PlanRas described [in,;.the;FSAR;;Which11sicontrolledl by 10CFR50154a)Ror)3)!to; prograssithat;are; controlled .;via the Administrative; Controls sectionLoflthejaproved TSLi The21dentificationf ofithel specific;11censeelcontrolled document *containing;thi.siroquirement;1siprovided;jn Enclosure;3BlofithelconversionlsubmittaM'1 Compliance;with;the' relocated; requirements; W1111not1be affected;byithisfproposedfchange,toltheicurrantjTechnical SpecificationsEiThe tequirediperiodic;surve111ances;W1))

continueltolbe: performed;.tolensureithatll.initsion parameters 1areimaintainedA;;ThereforeErelocationlofGthese requi rementswilt have:no; impaction:; system;operabil ity;.ot the;maintenanceioficontrollediparame.tersfwithin;11mitsJ l

CPSES Description of Changes - CTS 3N.1 6f 7/29/98

l TECHNICAL SPECIFICATION SCREENING FORM

! (1) TECHNICAL SPECIFICATION 3.1.2.5 BORATED WATER SOURCE - SHUTDOWN Applicable MODES: MODES 5 and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA j_ ls the Technical Specification applicable to:

l YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the l control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS.

if the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document.

(3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. Equipment required to perform this function includes, depending on operating conditions, a combination of; (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power source from operable diesel generators.

The boration subsystem of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to SDM. To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head.

The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational Attachment 21 9

occurrancts. Th3 SDM dzfin s ths d: gram of subcriticality that would ba obtaintd immediat:ly following ths ins:rtion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements ara met by means of adjustments to the RCS boron concentration.

The boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2.

For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3.

The boration system function at shutdown is used to provide reactivity control, in particular for dilution events. As part of the CPSES Safety Monitor shutdown model development, the issue of dilution events as an initiator to boiling or core damage was investigated. This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other p', ant shutdown models, including the Outage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is conciuded to be an insignificant contributor to shutdown risk for CPSES.

Given the foregoing, it is concluded th'at the boration flow path SSCs are shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

2. The Technical Specification may be relocated to a licensee controlled document.

Attachment 10

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-12 APPLICABILITY: CP, DC REQUEST: CTS 3.1.2.5 Borated Water Source - Operating (Comanche Peak & Diablo Canyon) DOC 11-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DOC. FLOG Response: DOC 11-01-R is revised and Technical Specification Screening Form for CTS 3.1.2.6 prepared to provide additionaljustification for the relocation. This justification shows that the boric acid storage system is not assumed to operate to provide negative reactivity control to mitigate any accident. The response for a boron dilution event would be to secure appropriate valves in the reactor makeup system. The SDM requirements provide sufficient reactivity margin to mitigate any anticipated event. S3nce the system does not mitigate an accident, there is no installed instrumentation which is used to detect or indicate a significant  : degradation of the RCS boundary. This system is also not associated with any variable, design l feature, or operating limit that is the initial conditions of an event which challenges a fission product boundary. This subsystem in the boration system is not a part of, nor does it support a system requiring that support to function as part of the success path to mitigate a design bases accident. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998. ATTACHED PAGES: Enci 3A 7 Att 21 11 & 12 l l l t

CHANGE NUMBER EitLG DESCRIPTION 11-01 R The CTS 3.1.2.6. Operating Borated Water Source and associated SR 4.1.2.6 are relocated to a licensee controlled document. This is consistent with NUREG- l 1431. The:boratedjwater;sourpelista?componentiwjthin & 3.1 12 the18 oration:Systen U The;functjon;of1the'Boration System;isLtolensurelthat negative; reactivity;controllis ayajl able;during[eachinode"ofi facil ity; operatj on d i Thi s! proposed [TS;reij sion; relocates; requirements;Ewhichido j potineetithe JS 'priteria $1ni10CFR5046,(c) (2) (11) tito , documents;with;establ i shed:.controlf programs GThi s regulation [ addresses'thelscopeland purposelofiTSr3In doing[so;11tisets!forth a specific; set;of; objective criteria ;forJdetermining;whichiregul atory; requirements [and operating " restrictions;should :belinc] uded:lin ;.the;TS7 ] j Rel ocationfof!these; requirements;allowsithe;TS' to; be reserved 1onlylfotithose; conditions;orilimitationsfupon ] reactotf.operatjonJwhichlarenecessarystoichviateithe possibility offanlabnormalisituationTorlevent giVing rise to2ap2jemediate ;threatitoithe; pub]ic; heal thland; safety therebyifocusi_ng ithelscope.: oftthe!TSEAnleval uationlofs the app 11cabilitylofLtheseicriteri attolthi s;speci ficati.onlis provjdedlin: Attachment?21I To~ ensure;an[ appropriate ] eye];of; control;1these requirements;wilEbelrelocatedLto1)?documentsLthatiare subject 1tolthelprovisionsiofjl01CFRl50j5942)Jother licensee;documentsIwhich have:similararegulatory1 controls  ! (e;g.' Tithe Quality;A.ssurance; Plan!;asidescribedLinf the ESAR;9hichus;contro11ed ; byi10CFR50: 54a)Zor; 3)ito { Programs;thatrare~contro11 ediviai the: Admini strati ve Control sisection~ofAthetimprovedlTS ?5Hheij identification of;thef speci fic ;] icenseelcontroll ed ; document Econtaining thi s; requirements; provided;1p; Encl osure L38: o.fithe c'onyersionisubojttalgi l C.ompliancej withithe;relocatedirequi rements;willinotibe affected1bylthis proposedichange:tolthe:currentiTechnical Specifjcat16ns M Ihe:requiredfperiodictsurve111ancesiw11] continue;to;be; performed;to; ensure!that; limits;on parameters 1arelmaintained; iThereforeitrelocati.oniof;these requirementslwill]hayeino';1mpactionisystem operability;oC q thejmaintenance oficontrolled;parametersf.withinflimits'l l 1 1 l CPSES Description of Changes - CTS 3M.] 7 7/29/98 J

4 TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.6 BORATED WATER SOURCES - OPERATING Applicable MODES: MODES 1,2,3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condTion of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _. 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of ti.e above questions is *NO", the TS may be relocated to a controlled document. (3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The equipment required to perform this function includes, depending upon operating conditions, combinations of (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric a::id transfer pumps, and (5) an emergency power supply from operable diesel generators. The boration subsystern of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain SDM. To accomplish this functional requirement, the boretion systems TS require a source of borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head. l The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient, in the case of a malfunction of the CVCS, which causes a boron l dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the 130'ation subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactWity with boron is not credited in the accident artalysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational Attachment 21 11

occurrine:s. Th3 SDM d; finis th3 degree of subcriticality that would b3 obtained immediat:ly following ths ins rtion or scram of all shutdown and control rods, assuming that the single rod assembly of hi t, hest worth is fully withdrawn. During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. Based on the foregoing, the boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1. The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem is not a design feature required to be operable to mitigate these events, and this TS does not satisfy criterion 2. i For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of th6 primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST and associated flowpaths is required as part of the ECCS TS. For the MSLB event, the sequence of events takes the plant to cold shutdown conditions and; therefore, boration of the RCS is necessary. However, the boration flowpath in this case is required as part of the ECCS function. The boration function of CVCS at operation is explicitly modeled in the CPSES Individual Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE model of this func%n includes both the boration path from the boric acid transfer tanks via the boric acid trancfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given the low failure probability and the redundancy and diversity of these flow paths, based on the results of the CPSES IPE. In addition, the ECCS function of the Refueling Water Storage Tank is explicitly modeled ir. the CPSES IPE and is being retained in the CPSES ECCS Tech Spec. Because the ECCS function of injection from the refueling water storage tank via the CCPs, and therciore one of the emergency boration flow paths, remains part of the improved tech spec, the emergency boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION This Technical Specifice Un is retained.

2. The Technical Specification may be relocated to a licensee controlled document.

l Attachment 21 12 L. A

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-13 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.1.4 Rod Group AUgr: ment Limits (Comanche Peak) DOC 12-07-A ITS 3.1.4 Bases Comment: The DOC states, for Required Action B.2.6, that "the ITS Bases discuss the accident analysis affected by rod misalignment." The associated Bases do not list the accident analyses that require re-evaluation, similar to that provided by the other Four Loop Group plants. List in the Bases the accident analyses that require re-evaluation. FLOG Response: The APPLICABLE SAFETY ANALYSES section of ITS 3.1.4 Bases provides an appropriate description of the various manners in which a misaligned rod can affect the safety analyses. The requirement in ITS 3.1.4 REQUIRED ACTION B.2.6 is to evaluate the safety analyses; the affected analyses are described more fully by the APPLICABLE SAFETY ANALYSES (ASA) than by the list transported from the CTS. In fact, many of the analyses listed (e.g., Decrease in Reactor Coolant Inventory in FSAR Section 15.6) are not affected by reasonable rod misalignments; whereas some transients that are sensitive to misaligned rods (most of the Power Distribution and Reactivity Anomaly accidents described in FSAR Section 15.4) are not listed. Because of the potential conflicts between the APPLICABLE SAFETY ANALYSES section and the list of CTS Table 3.1-1, it is preferable to not add the list, but refer to the APPLICABLE SAFETY ANALYSES section. The ITS Bases ACTIONS B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 are revised to indicate that the accident analysis presented in USAR Chapter 15 that may be adversely affected will be evaluated to ensure that the analyses results remain valid for the duration of continued operation. Callaway, Wolf Creek and Diablo Canyon have reviewed this Comment and concur with the above discussion. Their ITS Bases have been revised to delete the list of accident analyses that require re-evaluation and refer to FSAR Chapter 15. ATTACHED PAGES: Enct 58 B3.1-28

Rod Group Alignment Limits B 3-1-5 3.1 4 BASES ACTIONS determine that core limits will not be exceeded during a Design (continued) Basis Event for the duration of operation under these conditions. The accidentfanalyses presented;1n FSAR; Chapter 15((Reff 3)1thatTmaylbeladversely.affected#111:be evaluated 0 3.1 13 to' ensure;.that;thelanalyses;results; remain.ValidEfoNthe duration of. continued speration.under thesefconditionsMiA Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis. B-1-C .' 1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems. C.1.1 cr,d C.I.2 D.Til"and D.1:'2 More than one control rod becoming misal hned from its group average demand l position is not expected, and has the potential to reduce SDM. Therefore SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonabic, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the beric acid requ1Ced pumps. Boration will continue until the required SDM is restored. I , C-E-D:2 If more than one rod is found to be misa11gned or becomes l misaligned because of bank movement, the unit conditions fall l outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. l l (continued) CPSESMark-up ofNUREG-1,131 Bases - ITS 3.1 B 3.1-28 7/29/97 L _ __ _ ____ _ __ - _ - __-______-_____ _ __ _ __ _

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-15 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.1.4 Rod Group Alignment Limits CTS 3/4.1.3 Movable Control Assemblies (All FLOG Plants) DOC 12-14-M Comment: The ITS has changed the wording of the TS from "trippability" to " operability," and references TSTF-107 which is not yet approved (though it is expected to be approved with the OGs next revision of TSTF-107. The result is that the FLOG plants have inconsistently incorporated generic changes into the Bases (i.e., the Bases paragraphs for B.2.1.1 and B.2.1.2). This change is a less restrictive change in that it precludes LCO 3.0.3 entry for unforeseen inoperabilities. TSTF-107 needs to be discussed! approved at the next TSTF OG/NRC Meeting, and the FLOG will then need to incorporate the resulting generic TS requ%ments. FLOG Response: It is the FLOG's understanding that EXCEL Services Corporation met with the NRC on May 23,1998, to discuss TSTF-107. The result of that meeting has been reported to be agreement to approve TSTF-107 with a minor Bases change. Revision 1 of TSTF-107 has been incorporated into the FLOG submittals. In the ITS, rod operability is addressed in the Bases as trippability within the drop time requirements of ITS SR 3.1.4.3. If not met, Condition A would be entered which requires SDM verification and shutdown to Mode 3 in 6 hours, which then exits the LCO. In the CTS, the action for an untrippable rod is essentially the same as the ITS. No action is provided in the CTS for discovering in Mode 1 or 2 that a rod would not meet insertion time l requirements; therefore, CTS LCO 3.0.3 would be eritered. LCO 3.0.3 allows one hour to initiate ' a shutdown and 6 additional hours to reach Mode 3. Because the ITS only allows 6 hours to reach Mode 3 (instead of up to 7 as allowed by LCO 3.0.3), the change from "untrippable" to l

                                                                            " inoperable" in CTS 3.1.3.1 is more restrictive.

ATTACHED PAGES: EnciSA Traveler page Enci5B B3.1-27 1 I 1 I

                                                                                                                              .-                                            J

Industry Travelers Applicable to Section 3.1 1 TRAVELER # STATUS DIFFERENCE # COMMENTS l TSTF 9. Incorporated 3.1 1 NRC approved. Revision 1 TSTF-10 Not incorporated NA NRC rcjected- '( P 3.1 002 Revision 1 TSTT-11 Not incorporated -NA NRC rcjected. (P 3.1 002 Revision i TSTF 12 Incorporated 3.1 15 NRC approved. ITS Special Revision 1 Test Exception 3.1.10 is retained and re numbered as 3.1.8, consistent with this traveler and TSTF 136. TSTF-13. Incorporated 3.1 4 NRC approved. Revision 1 TSTF 14 Incorporated 3.1 13 NRC approved. .sith M000 2 (P 3.1 002 Revision --34 retaincd in ITS 3.1.0 A m m 1 J . L J 1 J 4. i , nVV = ''J-1R 3.1005 l TSTF 15 Incorporated NA NRC.; approved;;..Not yct approve jcp.3,1 002 Revision 1 gnerically. TSTF 89 Incorporated 3.1 8 NRC approved.

                                                                                       ^

T Incorporated 3.1 6 NRC requested changes. [ (,.3.1 15 ) evisi -~ - ( , TSTF-108 Net Incorporated NA Net NRC approved. es-ef ,g.3,1.ooi Rev.ision:1 3.1 21 traveler cutoff date. TSTF 110 Incorporated 3.1-10 NRC approved l 1g.3,1.oo4 Revision.2.4 TSTF to enhance travcler justification per NRC ( p.3.1 002 ccccents. TSTF-136 Incorporated 3.1-9,;3.1-15 cp.3,1 002 TSTF-141 Not incorporated NA Disagree with change; traveler issued after cut off date. I i I s

i Rod Group Alignment Limits B 3--l-5 3,1A l j BASES l ACTIONS In many cases, realigning the remainder of the group to the (continued) misaligned rod may not be desirable. For example, realigning I control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D mne be fully inserted and control bank C must be d in . insert approximately 100 to 115 steps. C_ w-0 4 1 15 Poweropera[maycontinuewithoneRCCA ERABLEi(1,en trippabl but misaligned, provided that SDM 1s verified within 1 hour. The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, ' aligning and starting the necessary systems and components to faitiate boration. Itiislassumed;that boration will(continue unt111SDHlequirementslareinet; B.2.2. B.2.3. B.2.4. B.2.5. and B.2.6 For continued operation with a misaligned rod, RTP-reactorfpower must be reduced, SDM must periodically be verified within limits, hot channel factors (Fo(Z) and Flu) must be verified within limits, and the safety aralyses must be re-evaluated to confirm continued operation is rarmissible. Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 3M. The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. When a rod is known to be misaligned, there is a potential to l impact the SDM. Since the core conditions can change with time,  ! periodic verification of SDM is required. A Frequency of i 12 hours is sufficient to ensure this requirement continues to be met. Verifying that Fa(Z)Rasiapproximated;by Ff(Z)TandFW(Z).c a and F1, are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power i distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate Fa(Z) and Fin. Once current conditions have been verified acceptable, time is available to perform evaluations of the affectedl accident analysis to (continued) CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-27 7/29/97 j L--___----____--_._-_-.__-____---

                 --                                                                                                                                 /

I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-16 APPLICABILITY: CA, CP, DC, WC . 1 REQUEST: ITS 3.1.4 Rod Group Alignment Limits (All FLOG Plants) ITS 3.1.4 Bases Generic Changes Comment: Generic Bases changes need to be discussed / justified. For example, the Bases Backgrcund discussion on the DRPI system has been revised and needs to be explained. FLOG Response: As discussed during a telecon with NRC Staff on June 25,1998, the scope of this RAI will be limited to the DRPI Background Bases. Changes fallinto one of four categories:

1. Specification re-numbering;.

2. Inclusion or shutdown rods;

3. Addition of plant-specific design information (e.g., number of control banks and shutdown banks);

4. Editorial corrections (e.g., the correct title for GDC-26)

5. Changes to the last paragraph.

Changes to the last paragraph were made since it was felt that this text went beyond the level of detail n. aired for the ITS Bases. Coil spacing dimensions are not critical to operator understr. Jing of this system. In addition, statements in the last paragraph in the ISTS concerning position indication accuracies are incorrect J ATTACHED PAGES: None 1 l l l i I

1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-17 APPLICABILITY: CP REQUEST: ITS 3.1.4 Rod Group Alignment Limits (Comanche Peak) DOC 12-19-LS-18 CTS SR 4.1.3.1.2 (rod motion SR frequency) ITS SR 3.1.4.2 Comment: Justification for changing SR frequency must be based upon plant specific reasons (i.e., on operating experience), and not solely on consistency with the STS. Provide adequate justification for changing frequency of control rod motion SR from 31 to 92 days. FLOG Response: DOC 12-19-LS-18 has been revised to provide additional justification for changing frequency of control rod motion SR from 31 to 92 days. ATTACHED PAGES: Enci 3A 10 l l

                                                                                                                                                                  \

L _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _

CHANGE Nl#EER HSBC DESCRIPTION the public health and safety. Therefore, moving TR.3.1 004 this detail is acceptable and is consistent with traveler TSTF 110. k;. 1. l 12 17 A Editorial changes made for clarity. Untrippable rods are addressed through Action a; hence, there is not additional need to exclude those rods from these required actions. 12-18 LG The technical contents of the Action Statement which allows continued power operation with a misaligned rod are moved to the Bases for ITS LC0 3.1.4, Action B.1. 12 19 LS 18 Consistent with NUREG 1431, Rev. 1, the frequency at which the rod motion surveillance is performed is _

                                                                                                                                                                        ^

exteng31 to 92,dA pmposed change;would M

                                                                      ~ . ,. -%them .p ng J mquency!.at which.eachirod cluster;controllassemb11Li slexercisoditoldemonstrate thelabiljty;ofitheirodsito; be stripped ti Current ; Technical Specification;Surve111ance Requirements [4;1:3;1'.2]

exercisesleachfrodicluster.;controUassemblyf at21eastiten steps;to; demonstrate:thef abiljty;of;theTr.odsito.be trjpped 27Thisisurve111ance; requirementiis1performediat legst once perf31?;daysf.Q; Verifying aachicontrolfggdli) OPERA 8LEwouldLrequirethat;eachirod;be;trippedqHowever! in:N00EST1:and 25 tripping,each;controlfod:Wouldiresult jW.radjatorlaxialfpowertilts;;orfoscillationsa Exercisingiesch individual;controlirod;every,92 days provides11ncreased; confidence ;that :all2 rods Lcontinueltoibe OPERABLE;withoutLexceeding.the[ alignment:11mit; eventif they; are:not[ regularly; tripped,EMoving ;each ?contro1Lrod byjl0 stepsiw11Enot';cause radia11orlaxiaEpowert t11ts, Tor osc111ations;Ltoloccur M The:92;dayJrequency; takes 11nto considerationfotherninformation;availableito theLoperator in athe; control; room end SRl3:1 4;1L which;isl performed morel.ffeguentlyland laddst to;the; determination Lof OPERABILI_TYJ of;thel rods M;Between;; ori dur:ing; required performances;of;SRJ3:1.4.2 (determination:of; control rod

OPERABILITYiby movement)",tifialcontrol' rod (s);;is l discovered;to belismovablesbutirenainsitrippablelithe l controllrod(s);is; considered;to be;0PERABLE;until'the l surveillancejinterval;expiresC" At

4ny,timeGiffalcontrol rod (sR1sLimmovableEa;determinationLotthaltrippability (OPERABILITY)1ofjthe: control; rod (s):must be madeRand appropriate: action;takenEThe~decreasesin the; testing frequencyito:stileast;once:per;92;daysfjs1 recommended b Generic; Letter;93 05'" Item _4,2,* "Contro1T Rod ~ Movement T_est.

CPSES Description of Changes - CTS 3N.1 10 72988 _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ ____O

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-18 APPLICABILITY: CP REQUEST: ITS 3.1.4 Rod Group Alignment Limits (Comanche Peak) DOC 12-20-A CTS SR 4.1.3.1.3 (rod drop time SR frequency) , ITS SR 3.1.4.3 Comment: In DOC 12-20-A the ITS SR 3.1.4.3.is incorrectly referred to as SR 3.1.5.3. Correct DOC. FLOG Response: DOC 12-20-A has been revised to reference ITS SR 3.1.4.3 instead of SR 3.1.5.3. ATTACHED PAGES: EncI 3A 10a I t I. L____-________ _ .1

CHANGE NUMBER EE DESCRIPTION 12 20 A The Action Statement in the current TS to restore the rod drop time 16 within limits as a condition for MODE 2 is captured in frequency for the performance of ITS SR 3.114:3 3.1.5.3 . 0 3.1-18 12 21 Not used. 12 22 M This change, in accordance with NUREG 1431 Rev. 1, provides a new ACTION in the event the allowed outage times are not met for the rod misalignment actions. Prior to this change LC0 3.0.3 would have been entered allowing for 1 hour prior to placing the plant in HOT STANDBY within the next 6 hours. This change is more restrictive in that the 1 hour time frame is eliminated. 13 01 LG Consistent with NUREG 1431, Rev. 1, the operability attributes of equipment and components are described in the Bases. The proposed elimination of the accuracy attributes of the Digital Rod Position Indication System and Demand Position Indication System from the Specification on positien indicating systems would have no impact on OPERABILITY of these systems because the design of these systems is fixed. Furthermore. ITS LC0 3.1.4 requires that all individual indicated rod positions be within 12 steps of their group step counter demand position. Therefore, the LC0 effectively establishes the accuracy requirements for the rod position indicating system, and eliminating it from the specification for indicating systems would have no effect on the OPERABILITY j of the indicating systems. J l CPSES Description of Changes - CTS 3N.] 10a 72 9/98 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-_Y

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-20 APPLICABILITY: CP,DC l REQUEST: ITS 3.1.7 Rod Position Indication CTS 3.1.3.2 Position Indication Systems - Operating (Comanche Peak & Diablo Canyon) DOC 13-08-LS-20 & 13-09-LS-23 & 13 06-A JFD 3.1-7 & 3.1-12 Comment: The ITS adopts Conditions and associated Required Actions from the Callaway's CTS, addressing more than one inoperable digital rod position indicator (DRPI) per group, which is not addressed in either the STS or the CTS. Furthermore, not all associated CTS Required Actions have been retained in the ITS; the Required Actions to take manual control of the rods and to record reactor coolant temperature every hour have not been retained. These actions, in one case affect rod movement and in the other case provide an indication that the rod (s) position may have changed, and therefore have a bearing on SDM and therefore should not be deleted if the Callaway condition of more than DRPI per group inoperable is retained. Either retain the CTS requirements and adopt the STS requirements, or provide a better justification for the ITS proposals of adopting the Callaway CTS requirements. This change is based upon proposed change WOG-73, Rev 1; which eventually may become a TSTF change request. What is the status of WOG-73, Rev 17 The STS wording of the note permitting separate condition entry should be retained with the STS Conditions and Required Actions. FLOG Response: No change. DCPP and CPSES wish to continue pursuing this change pending NRC review of WOG-73, Rev.1, which has been submitted to the NRC as TSTF-234. This proposed change provides a specific set of compensatory actions and a 24 hour AOT for a situation previously unaddressed. Thus, the requirements of LCO 3.0.3 would be applicable and a plant shutdown would be required within 1 hour. However, the partialloss of RCCA position indication is not an assumed initiating event, nor does it effect the outcome of any analyzed accident. Therefore, an one hour shutdown response l

is considered unnecessary when compared to other situations of similar severity where guidance was provided.

The contribution to plant safety of a well-balanced approach in avoiding unnecessary shutdowns is sufficient to warrant prompt NRC review and approval of TSTF-234. It would also be sufficient ' to warrant a separate License Amendment Request were the TSTF to be withdrawn from this submittal. l l ATTACHED PAGES: l l None 1 . i

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-21 APPLICABILITY: CP, DC REQUEST: CTS 3.1.3.3 Position Indication Systems - Shutdown (Comanche Peak & Diablo Canyon) DOC 14-01-R Comment: The DOC needs to specify where the CTS specification is being relocated. Correct the DOC. A relocated screening form is not provided for this relocated specification. FLOG Response: For CPSES, DOC 14-01-R has also been revised to provide additional justification for the relocation and to reference new Attachment 21 containing relocation screening forms. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998. For Diablo Canyon, License Amendment (LA) 120/118 (dated 2/3/98) eliminates this RAI. LA 120/118 relocated ten TSs in accordance with 10 CFR 50.36. Thus, CTS 3.1.3.3 no longer exists and DOC 14-01-R is not applicable to DCPP. CTS 3.1.3.3 was relocated to ECG 41.1 which is part of an NRC approved program controlled by 10CFR50.59 and discussed in FSAR Chapter 16. ATTACHED PAGES: Enci 3A 12 & 12a Att 21 13 & 14 i f I { l l l l l L }

CHANGE NUMBER EBC DESCRIPTION 13 07 M The proposed modifications to the SR would require a verification of agreement between digital and demand indicator systems prior to criticality after each removal of the reactor vessel head, instead of every 12 hours. This reflects a reorganization of surveillance requirements in the ITS. The requirement for a 12 hour comparison would be moved to SR 3.1.4.1 in the ITS. The post vessel head removal requirement would be a new specification that demonstrates rod position system OPERABILITY based on a comparison of indicating systems. The Frequency requirement of prior to criticality after each removal of the reactor vessel head would permit this comparison to be performed only during plant outages that involve plant evolutions (vessel head removal) that could affect the OPERABILITY of the rod position indication systems. The Frequency change is based on traveler TSTF-89. 13 08 LS 20 Adds a provision, from Callaway's current TS. which would, under certain conditions, allow continued operation with more than one inoperable DRPI per group. A separate l condition entry allowance is permitted for each inoperable rod position indicator and each demand position indicator. 13 09 LS 23 Not applicable to CPSES. See Conversion Comparison Table (Enclosure 3B). , 14 01 R Shutdown Position Indication System specification 3.1.3.3 is relocated to a licensee controlled document. ! This is consistent with NUREG 1431 Rev. 1. Thell0jgital 0 3.1 21 Rodfosition:IndicationfSystealis;usediforJverification offagreement.iwithdemandedposition? ! l l ThisiproposedTSirevisjonj relocates" requirements;T which;do notimeetithe;TSJeriteria;inl10CFR50.36(c)(2)(11)Rto i documents;with :establi shedicontroliprograms QThi s regul ation; addresses;the; scope [and:purposef of1TS . iip  ! doing!soRit;setsiforth alspecificsset otobject1W  ! criterialforZdeterminjng which; regulatory 1 requirements;and operating;restrjetjonsishouldibelincludedlin1the1TS7 l Relocation _ofithese;requirementsiallowsithe;JSitobe  : reservedlonlylforlthose:conditionsior limitations;upon reactor { operation:which X e nece.ssary;to;obvjate;the possibility offan~abnormalisituation:or; event;givi.ng; rise to [an11mmediateithreat;tosthel public1 health ~and; safety I thereby;focusjpgithefscope[ofathekTSEAn; evaluation;off the I appl icabil ity;otthese: crjterj a ;to;thi s;speci fication ii s ) providedLin: Attachments 217 CPSES Description of Changes - CTS 3N.] 12 7/2988 j

i CHANGE NUMBER NSliC DESCRIPTION

                                                                                                            -                             -                  N          .

ensurelan approptiatellevellof; control,ithese 0 3.1 21 tequi.tementsLwill;belrelocateds tollEdocuments .thatfare I ' l subjectito.:the~provisionsiofi10 CFRl50;59lJ2)Lothet 11censee; documents 1whichfhaveisimilat regulatoryLcontrols (e;g?&the Quality Assurance l Plan,Las.describediin the ESARJwhichll,is[ controlled byl10CFR50.54a),Eot 3){to Prograssithat articontrolled;via the1 Administrative Control s1section iofithe Limprov.edjTS OThe ; identi fication l ofithe!specificl1icenseelcontrolled; document containing i thisfequirementJs;provided!1n

Enclosure:

3B;of;the conversion submitta1 G i Compliance withithelrelocated; requirements:willinat:be i Affectedlby this proposed changelto theicurrentJechnical I f Specifications.HIherequirediperiodicjsurve.illances;will continue (to; helperformedito ; ensure 1that211mit$ con parameters are.,maintajnedCJherefore?.LrelocationfoLthese requi rements; will ihaveino ' impact ;on " system ;operabil ity 1or the:maintenancefof controlled' parameter.s;withinilimits!  ! 15 01 R Not used;ar y ma .c to CPSES. Sec Conversion DC ALL-004 1 Cc;;;parison Tabic (Enclosurc 30). i 15-02 A The Rod Drop Time SR 4.1.3.4.a is moved to the Control Rod ITS LC0 3.1.4, as SR 3.1.4.3. This is consistent with NUREG 1431. l 16 01 LS 14 This TS would be revised to apply to shutdown " banks" l instead of shutdown " rods": this is consistent with NUREG-1431 Rev. 1. The current Action Statement permits one l rod to be inserted beyond the limits: the proposed ITS CONDITION A would allow one or more banks to be inserted beyond the limit. 16-02 H The proposed changes to the Action Statement wo21d require that the shutdown banks be aligned within limit s and that SDM be verified or restored. The new Action Stn ement l \ CPSES Description of Changes - CTS 3N.1 12a 7/2988 l

TECHNICAL SPECIFICATION SCREENING FORM 1 1 (1) TECHNICAL SPECIFICATION 3.1.3.3 POSITION INDICATION SYSTEM - SHUTDOWN I l Applicable MODES * " 3,4 AND 5 (* With the reactor trip breakers in the closed position. See Special Test Exceptions Specification 3.10.5.) (2) EVALUATION OF POLICY STATEMENT CRITERIA  ! Is the Technical Specification cpplicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. I _ X (2) A process variable, design feature, or operating restriction that is an I initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.  ;

                                           ,_,,        1         (3)      A structure, system, or component (SSC) that is part of the primary         1 success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.                               )

I _ 1 (4) An SSC which operating experience or probabilistic safety assessment l (PSA) has shown to be significant to public health and safety. j l If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. j l if the answer to all four of the above questions is "NO", the TS may be relocated to a controlled j document. j (3) DISCUSSION The Digital Rod Position indication System verifies the agreement with demanded position at specific steps within the rod position range. l l The Digital Rod Position Indication System is not an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either

                                                                                                                                                    )

l assumes the failure of or presents a challenge to the integrity of a fission product . barrier. The Dig:tal Rod Position indication System TS is not applicable to installed instrumentation that is used to detect and indicate in the control room a significant i abnormal degradation of the RCPB. Therefore, this TS does not satisfy criterion 1.  ! The position indication system TS, for shutdown conditions, is not associated with a l process variable, design feature, or operating restriction that is an initial condition of a J DBA or transient analysis tnat either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the TS does not satisfy criterion 2. The Digital Rod Position Indication System TS does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or Attachment 21 13

transient thit eithir assum:s the frilura of or prssints a challings to tha integrity of a fission product birrier. Th re fore, ths TS does not srtisfy crit:rion 3. l From Ref. 3, the Digital Rod Position Indication System has not been shown to be a l significant risk contributor to public health and safety by either operational experience or PSA. The Digital Rod Position Indication System is not modeled in the CPSES iPE. This TS does not satisfy criterion 4. t I

                                                                                                                                                                                                .l3 (4)      CONCi.USION                                                                                                                                                           '

l- ,_ This Technical Specification is retained. l I 1 The Technical Specification may be relocated to a licensee controlled document. l I~ l: i l l l t , l i I Attachment 21 i 14

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-24 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.1.4 Rod Group Alignment Limits (All FLOG Plants) JFD 3.1-5 & 3.1-6 Comment: Rewording of LCO and Condition A approved, contingent upon OG resubmittal of change request TSTF-107 (revision) as discussed with TSTF. FLOG Response: See the response to Comment Number 3.1-15. The FLOG has incorporated TSTF-107, Revision 1. ATTACHED PAGES: None l 1 l 1 i l l I

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O3.1-25 APPLICABILITY: CA, CP, DC, WC q REQUEST: ITS 3.1.4 Rod Group Alignment L!mits (Ali FLOG Plants) JFD 3.1-16 Comment: Inclusion of SR 3.2.1.2 to Required Action B.2.4 is approved; ensure OG submit WOG-105 as a TSTF change request. I FLOG Response: At the June 23-24,1998 meeting of the Westinghouse Owners Group MERITS Mini-Group, traveler WOG-105 was discussed. The remaining action on this traveler was assigned to Westinghouse to expand this change to also apply to ISTS 3.2.1 A, "Fa (Z) (F,y Methodology)." However, this additional work has no impact on the manner in which the FLOG l has incorporated this traveler's additional restriction. The TSTF will be submitted to NRC expeditiously. ATTACHED PAGES: None _ _ - - - _ _ - - _ _ _ _ _ _ _ _ _ _ - _ . - - - . _ _ _ E

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-26 APPLICABILITY: CP REQUEST: ITS 3.1.8 Physics Tests Exceptions - Mode 2 (Comanche Peak) JFD 3.1-20 Comment: The Conversion Comparison Table Indicates that this change is applicable to Oomanche Peak, however, the STS mark-up does not have it included. Include change in STS mark-up and in ITS. FLOG Response: The ITS markup has been revised to include JFD 3.1-20. ATTACHED PAGES: Encl SA 3.1-21 & 3.1-22 Enci5B B 3.1-54,55 & 56 l

                                                                                                                                                                                                                                            )

( l

3.1 REACTIVITY CONTROL SYSTEMS 3.1.10 3.1;8 PHYSICS TESTS Exceptions-MODE 2 i LC0 3.1.10 3;1.8 During the performance of PHYSICS TESTS, the requirements of LC0 3.1.43, " Moderator Temperature Coefficient (MTC)"; 3.1-9I LC0 3.1.54. " Rod Group Alignment Limits": LC0 3.1.65, " Shutdown Bank Insertion Limits": LC0 3.1.76, " Control Bank Insertion Limits"; and LC0 3.4.2, "RCS Minimum Temperature for Criticality" may be suspended, provided:

                                                                                                                                                                                              >    ;B.PS3 -

D -

a. RCS lowest perating loop average temperature is 2-E5313 3,1 20 541 F: and -- --

o.3.1 26

b. SDM is 2-tid]! ok/k. Within;.the limits provided;in;the '3.1-1: 1 COLR;;and

c. THERHAL;/ t 0WER isli 5t RTP2 t 3.1 APPLICABILITY: MODE 2 during PHYSICS TESTS.

ACTIONS

                                                                                                                                                                                            =

CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to 15 minutes restore SDM to within limit. AND i l A.2 Suspend PHYSICS TESTS 1 hour exceptions. l B. THERMAL POWER not within B.1 Open reactor trip breakers. Immediately limit. , t (continued) CPSESMark-up ofNUREG-14.si - ITS 3.1 3.1-21 7/19/98

l PHYSICS TESTS Exceptions-MODE 2 3.1-9 > 3.1.10 3;1'.8 3.1-15s ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCSlowesthrating[ C.1 Restore RCS lowest 15 minutes 3.1 20 loop average temperature N (operating]oop average 0 3.1-26 not within limit. temperature to within - limit. D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met. l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.10.13.118.1 Perform a CHANNEL OPERATIONAL TEST on W+tMn 3.1 21 power range and intermediate range 12 hours channels per SR 3.3,1;7;1SR;;3.3.1.8Eand pRriorto l Table 3,3.1 1. initiation 8 of PHYSICS TESTS I SR 3.1.10.2 3.1.8,2 Verify the RCS lowes h 30 minutes average temperature is z-fbatt 541*F. 3- [ 26

                                                                                                                                                                                                 .          {
                                                                                                                                                                                                  .B-PS-SR;3?l;8.3       Verify; THERMAL 7 POWER isR 5tLRTP:                                                           1 hour
                                                                                                                                                                                              ' 3.1-13 !

l I SR 3.1.10.3 Verify SDM is 2 1.5% ok/k within,thellimits 24 hours 3.1-l' j E13;1;8.42 provided;in the'COLR; CPSESMark-up ofNUREG-1431 - ITS3.1 3.1-22 7/19/98

i PHYSICS TESTS Exceptions VJDE 2 i B 3.1.-108

                                                                                                                                                                                                                                                             ~

l BASES meet Criteria 1, 2, and 3 of the NP,C Policy Stat;x nt 10;CFR 50;36(c)(2)(ii). l Reference 6 allows special test exceptions (STEs) to be included as part of the LC0 that they affect. It was decided, however, to retain this STE as a separate LC0 because it was less cumbersome and provided additional clarity. LCO This LC0 allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met. The requirements of LC0 3.1.34, 3.135, LC0 3.1.56. LC0 3.1.67 and LC0 3.4.2 may be suspended during the performance of PHYSICS TESTS provided:

a. RCS lowes operatin cloop average temperature is

t [531]541 l b. SDH is c [1.0]! s/k- within;the; limits specified;in_the,COLR; and cf THERMALLPOWERsisi5t RTP7 l APPLICABILITY This LCO is applicable in MODE 2 when performing low power PHYSICS TESTS. The applicable PHYSICS TESTS are performed in MODE 2 at HZP. I l Other PllYSICS TESTS are performd in "00E 1 and era addressed in i LC0 3.1.0, "PllYSICS TESTS Excepticr,s "00E 1. "

l i ! ) i i I ! ACTIONS A.1 and A.2 l

                                                                                                                                                                                                                                                               )

If the SDM requirement is not met, boration must be initiated j' promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDH is within limit. CPSESMark-up ofNUREG-1431 Bases - IIS 3.1 B 3.1-54 7/29/98 I

PHYSICS TESTS Exceptions-H0DE 2 B 3.1.108 , BASES l Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification. I IL1 When THERMAL POWER is > 5* RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor { beyond its design limits. Immediately opening the RTBs will shut down I the reactor and prevent operation of the reactor outside of its design limits.

                                                  ,          n            _

When the RCS lowes operating 300[T,, is < 531541*F the 0 3.1 26 appropriate action 1 . rcetcrc .., to within its specified limit. The allowed Completion Time of 15 minutes provides 1 time for restoring T., to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 531 5.41*F could violate the assumptions for accidents analyzed in the safety analyses. D.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a H0DE in which the ! requirement does not apply. To achieve this status, the plant must be l brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on ! operating experience, for reaching MODE 3 in an orderly manner and l without challenging plant systems. l 1 \ l l l l SURVEILLANCE SR 3.1.810.1 REQUIREMENTS The power range and intermediate range neutron detectors must be 1 l Verified to be OPERABLE in MODE 2 by LC0 3.3.1, " Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel TR-3'1-001 withir,12 hours prior to initiation of the PHYSICS TESTS. - This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS. The 12 hour ti;c limit 1 (continued) CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-SS 7/2988 3

                                                                                                                                                                                                                     )

PHYSICS TESTS Exceptions-MODE 2 B 3.1.108 BASES is :sfficicr.t to ca;;rc th:t the instr;; cat: tion i; OPEPJ.SLC sheetly Mforc initi ting P!lYSICS TESTS. m 3.1 001 SR 3.1308.2 - 3 Verification that the RCS lowest .ating31 p T.,, is a SM 0 3.1 26 541*F will ensure that the unit is mn. vym at'ing in a condition that could invalidate the safety analyses. Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated. SR:311.873 Verification 1thatJthejTHERMAL POER j_s]h5(iRTPJ111[ensuretthat;the plantiisinotloperatingfjnla conditionithat:coul.diinvalidatel;the safety ana]ysesEVerificationiof the1THERNAL e z POER[atsafrequency;ofslihour duri nglthe.) performance Loff,the; PHYSICS; TESTS 3v111:ensurelthat1the initiaUconditionsTofithelsafetyianalyses:arenotiviolated; 2 SR 3.1.408.43 l Veri fj cation;thatithe:SDHij s Lwithj ol imitsispeci fiedijn;the; COLR ensures 1that~,;fotthe; specific 1RCCA?and;RCS; temperature; manipulations performed l,during fHYSICSITESTSEthe l pl ant ;1 sinotioperatinglin ia condition;theticoul_dlinyalidate;the: safety 2analysi s: assumptions O The l SDHiverifjeationican;be; facilitated lthroughithe;uselofitables prepared by;the: core;designersliniwhichthelreactiv.ity: effects 2 expected;during thelPhysicsiTesting;have,been previously; considered? l The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration;

b. Shutdown ~and; Control bank position:

l

c. RCS average temperature:

d. Fuel burnup based on gross thermal energy generation:

e. Xenon concentration:

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-56 7/29/98 l __

1 I I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-27 APPLICABILITY: CA, CP, DC, WC i l REQUEST: ITS 3.1.1 Shutdown Margin (All FLOG Plants) JFD 3.1-18 Comment: This modification adds a Mode change restriction from Mode 6 to Mode 5, as discussed in CN 1-02-LS-1 of 3.0. The discussion provided is inadequate to evaluate the j necessity of the mode change restriction. In general, throughout the submittal, justifications for i notes prohibiting mode changes are inadequate. Provide explanations / justifications that present { specific conditions that would necessitate the note, j l FLOG Response: A Reviewer's Note in STS LCO 3.0.4 states: "LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all ) MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the 1 licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be lacluded in individual LCOs to justify this i change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate l NRC staff review of a conversion to the STS." Based on this Reviewer's Note, a matrix of this l evaluation was placed in the NSHC LS-1 in Enclosure 4 of the Section 3.0 package (Attachment l No. 6). I JFD 3.1-18 has been revised to incorporate additional justification from NSHC LS-1 from Enclosure 4 of the Section 3.0 package (Attachment No. 6). JFD 3.1-18 has been revised to l include: "LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or cther specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in' all MODES. ITS LCO 3.1.1 was modified by a Note stating: I "While this LCO is not met, entry into MODE 5 from MODE 6 is not permitted." Entering MODE 5 without SDM limits met implies that boron concentration in MODE 6 is not met. Under these conditions, a transition to MODE 5 should not be attempted until MODE 5 SDM limits are met. Inadvertent boron dilution events are precluded in MODE 6 via administrative controls [that close the dilution source valves), whereas dilution events are not [ physically precluded] in MODE 5. Therefore, the transition from MODE 6 to MODE 5 should not be allowed in the SDM initial I condition for a MODE 5 dilution event is not met." l \ l l ATTACHED PAGES: j l Encl.6A 3 and 3a I

CHANGE NUMBER JUSTIFICATION 3.1 11 Not Used. 3.1 12 The Required Actions for inoperable DRPI in ITS 3.1.7 are revised per the current licensing basis to note that the use of movable incore detectors for rod position verification is an indirect assessment at best. The position of some rods can not be ascertained by this method. 3.1 13 This change adds an LC0 requirement and SR to MODE 2 Physics Tests Exceptions 3.1.8 to verify that thermal power is less than or equal to 5 percent RTP. The LC0 requirement and SR were added to verify that THERMAL POWER is within the defined power level for MODE 2 during the performance of Physics Tests, since there is an Action that addresses THERMAL POWER not within limit yet there was no corresponding LC0 or surveillance requirement. The Surveillance Frequency of 1 hour is retained from the current TS. This change is based on TSTF 14. Revisi;n 3. TR 3.1 005 3.1-14 Not used. 3.1 15 Consistent with TSTF 12 Revision 1 ISTS LCOs 3.1.9 and 3.1.11 are deleted. The physics tests contained in LC0 3.1.9 were only contained in some initial plant startup testing programs. The physic test exception can be deleted since these physics tests are never performed during post refueling outages. The physics test that LC0 3.1.11 required was the Rod Worth Measurement in the N 1 condition. The use of other rod worth measurement techniques will maintain the shutdown margin during the entire measurement process and still provide the necessary physics data verification. Since i the N 1 measurement technique is no longer used, the SDM test exception can be deleted. This change and traveler TSTF-136 renumbers ISTS 3.1.10 to ITS 3.1.8. 3.1 16 This change adds the requirement to perform SR 3.2.1.2 in addition to SR 3.2.1.1 during performance of ITS 3.1.4 Required Action B.2.4. The intent of Required Action B.2.4 is to verify that Fa(Z) is within its limit. Fa(Z) is approximated by Fj(Z) (which is obtained via SR 3.2.1.1) and F#(Z) (which is obtained via SR ' 3.2.1.2). Thus, both Fj(Z) and Fl(Z) must be established to verify Fa(Z) . This change is consistent with traveler WOG-105. 3.1 17 Consistent with current TS LCO 3.1.3.2, ITS 3.1.7 Condition C is clarified to state that the inoperable position indicators are ! inoperable DRPIs. 3.1 18 A MODE change restriction has been added to ITS 3.1.1. in the LC0 applicability. per the matrix discussed in CN 102 LS 1 of the 3.0 oackana3TC0;3;0Whas;beenyevised;so;that changes L 0 3.i m ' _.iMODESlorf other; speci fi ed . condi tionsl i n' the tAppl icabil ity I that;are partfofla:shutdownLof the; unit?shall:not1be CPSESDifferencesfrom NUREG-1431 -ITS3.1 3 7/29/98

f CHANGE NUMBER JUSTIFICATION prevented.S;Inledditioni;LCO,3.A.4;hasbeenirevised,so.that 0 3'1 27 ittis:only; applicable forLentryLinto.::a z MODE'!orlothat specifiediconditionsl1n;the:Applicabi.1.1ty in; MODES;;17 2E3; end3E;Jhe MODE: change restri.ction.sjin,LCO 3.0.41werel:prevjously applicable; inia 11': MODES;2 ITS'LC013~1;r wasimodified byla; Note stating;While;this:LColi s;not;aet ;;; entry;jnto : MODE; 5;from.. MODE; 6 i s l noti permitted;"$ Entering; MODE ;51withoutlSDM ;11mits;; met timpl ies , that,boroniconcentrationjinLMODE;6ll.isinotTeetC1.Underithese l conditionsRaltransition;.to: MODE 51shouldinot';befattemptedLunt11 MODE 15l:SDM11mitslare:eet;:flInadvertent; boron;dilutionevents:are precludedI1n MODE.6lyiafadministrative; controls:[theticlose the dilutionisourceival ves] ; ;.whereas;dilutionieventsfare:not [ physically; precluded]iin MODE;5njThereforeEthe1 transition;fros l MODE:;6.;to; MODE 15)hou_1d;not; bejallowedlljnithe;SOMiinitial l condition lforja3J0DE;5;dilutioneventiis;not: set; CPSES Differencesfrom NUREG-1431 -ITS3.1 3a 7/29M8

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.1-28 APPLICABILITY: CA, CP, DC, WC REQUEST: Relocated Specifications (All FLOG Plants) Comment: Comanche Peak, Wolf Creek, and Callaway have not provided relocated screening evaluations / forms for any of their specifications relocated to licensee controlled documents. Diablo Canyon has not provided relocated screening forms for all of their specifications relocated to licensee controlled documents. Provide necessary relocation screening evaluations / forms. FLOG Response: All relocated specifications have been provided the necessary relocation screening evaluations / forms which are contained in Attachment 21. For Callaway and Wolf Creek, Section 3.1 specifications were previously relocated by Amendment No.103 and 89 respectively. Therefore, none of the relocation DOCS apply to Callaway and Wolf Creek and this question is not applicable to these plants. ATTACHED PAGES: Att 21 1 thru 14 1 i j h i \

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.1 BORATION FLOW PATHS - SHUTDOWN 1 Applicable MODES: MODES 5 and 6 l (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YE3 NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis i that either assumes the failure of or presentr a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primsry success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "Y ES", then the Technical Specification (TS) shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. l (3) DISCUSSION The Bases for this limiting condition for operation (LCO) state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The boration subsystem of the chemical and volume control system (CVCS) provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain shutdown margin (SDM). T*. accomplish this functional requirement, the boration systems TS r6 quire a source of borated water, one or more I flow paths to inject this borated water into the reactor coolant system (RCS), and appropriate charging pumps to provide the necessary charging head. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate j valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel 1 1 i

l d: sign limits will not be exceeded for normal shutdown cnd anticipated op: rational occurr:nces. Th3 SDM d: finis ths d: gree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. Based on the foregoing, the boration subsystem is not installed instrumentation that is used to detect or indicate a significant degradation of the reactor coolant pressure boundary (RCPB); therefore, this TS does not satisfy criterion 1. The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2. For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. The boration system function Et thutdown is used to provide reactivity control, in particular for dilution events. As part of the GPSES Safety Monitor shutdown model development, dilution events as an initiator to boiling or core damage were investigated. This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other plant shutdown models, including the Outage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is concluded to be an insignificant contributor to shutdown risk for CPSES. Given the foregoing, it is concluded that the boration flow path SSCs are shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION _ This Technical Specification is retained. K. The Technical Specification may be relocated to a licensee controlled document. l 1 l l

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Attachment 21 2 - ]

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.2 BORATION FLOW PATHS - OPERATING Applicable MODES: MODES 1,2,3 and 4* (*A maximum of two charging pumps shall be OPERABLE whenever the 16.nperature of one or more of the RCS cold legs is less than or equal to 350*F except when Specification 3.4.8.3 is not applicable.) (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: YES NO _ 1 (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barriei. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to i the integrity of a fission product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES", then the Technical Specification (TS) shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION The Bases for this limiting condition for operation (LCO) state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The boration subsystem of the chemical and volume control system (CVCS) provides l the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain shutdown margin (SDM). To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the reacto, coolant system (RCS), and appropriate charging pumps to provide the necessary chsijng head. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operatian of Attachment 21 3

th3 bor: tion subsyst m is not tssumed to mitigat3 this ev:nt. Furth rmors, Rif. 3 not:s thit thm normil capibility to control rectivity with boron is not credit d in tha accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requiremer,ts are met by means of adjustments to the RCS boron concentration. Based on the foregoing, the boration subsystem is not installed instrumentation that is used to detect or indicate a significant degradation of the reactor coolant pressure boundary (RCPB); therefore, this TS does not satisfy criterion 1. The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2. For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. The boration function of CVCS at operation is explicitly modeled in the CPSES Individual Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE model of this function includes both the boration path from the boric acid transfer tanks via the boric acid transfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given ( the low failure probability and the redundancy and diversity of these flow paths, based d on the results of the CPSES IPE. In addition, because the ECCS function of injection from the refueling water storage tank via the CCPs ( and therefore one ca the emergency boration flow paths) remains part of the improved tech specs in another section, the emergency boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION

             ,_       This Technical Specification is retained.

l

             .X.       The Technical Specification may be relocated to a licensee controlled document. 1 4
                                                                                                             )

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.3 CHARGING PUMPS - SHUTDOWN Applicable MODES: MODES 5 and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or Operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is pr;t of the primary success path and which functions or actuates to :nitigate a DBA or trans!ent that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES". then the TS shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. Equipment required to perform this function includes: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power source from operable diesel generators. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. Attachment 21 5 C_____.____ . _ _ _ _ . _ _ _ ~ . - . . - - - - - a

Th3 boration subsyst:m TS is not cpplicabl3 to install:d instrum2ntation used to d tect or indicats a significant degradation of ths RCPB; thirtfora, this TS does not satisfy criterion 1. The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2. For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a SSC that is part of the primary success path and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satis'y criterion 3. The boration system function at shutdown is used to provide reactivity control, in particular for dilution events. As part of the CPSES Safety Monitor shutdown model development, the issue of dilution events as an initiator to boiling or core damage was investigated. This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other plant shutdown models, including the Outage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is concluded to be an insignificant contributor to shutdown risk for CPSES. Given the foregoing, it is concluded that the boration flow path SSCs are shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION

                                                                 ._        This Technical Specification is retained.

_K. The Technical Specification may be relocated to a licensee controlled document. Attachment 21 j 6 i

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.4 CHARGING PUMPS - OPERATING Applicable MODES: MODES 1,2,3* and 4* " ("The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the charging pump declared inoperable pursuant to Specification 3.1.2.4 provided the charging pump is restored to OPERABLE status within 4 hours after entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375*F, whichever comes first.

                             **In MODE 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.)

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Techn: cal Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

              .__            1         (3)     A structure, system, or component (SSC) that is part of the primsry success path and which functions or actuahs to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the TG shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION l The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The equipment required to perform this function includes: (1) borated water sources, (2) charging pumes, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from operable diesel , generators. The boration subsystem of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain SDM. To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head. l l Attachment 21 l 7 L____________-_. _ _ _ - ]

Th3 boration subsyst:m is not Essumid to operata to mitigato tha consequences of a DBA or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is ! not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to l control reactivity with boron is not credited in the accident analysis. SDM requirements l provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be l exceeded for normal shutdown and anticipated operational occurrences. The SDM defines ) the degree of subcriticality that would be obtained immediately following the insertion or l scram of all shutdown and control rods, assuming that the single rod assembly of highest ! ' worth is fully withdrawn. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. Based on the foregoing, the boration subsystem is not installed instrumentation that is used i to detect or indicate a significant degradation of the RCPB; therefore, this TS does not ) satisfy criterion 1. I The boration subsystem TS is not associated with a process variable, design feature, or

operating restriction that is an initial condition of an event that assumes failure of or

challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2.

For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to estaolish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either I assumes the failure of or presents a challenge to the integrity G a fission product barrier; therefore, the TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the refueling water storage tank (RWST) and associated flowpaths is required as part of the emergency core cooling system (ECCS) TS. For the main steamline break (MSLB) event, the sequence of events takes the plant to cold shutdown conditions and; therefore, boration of the RCS is necessary. However, the boration flowpath in this case is required as part of the ECCS function. The boration function of CVCS at operation is explicitly modeled in the CPSES Individual Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE model of this function includes both the boration path from the boric acid transfer tanks via the boric acid transfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given the low failure probability and the redundancy and diversity of these flow paths, based on the results of the CPSES IPE. In addition, the ECCS function of the centrifugal charging pumps is explicitly modeled in the CPSES IPE and is being retained in the CPSES ECCS Tech Spec. Because the ECCS i function of injection from the refueling water storage tank via the CCPs, and therefore one of I the emergency boration flow paths, remains part of the improved tech spec, the emergency l boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy l criterion 4. (4) CONCLUSION _. This Technical Specification is retained. \

                .1        The Technical Specification may be relocated to a licensee controlled document.

( Attachment 21 1 8 e _ _ J

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.5 BORATED WATER SOURCE - SHUTDOWN Applicable MODES: MODES 5 and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: YES NO _ X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ X (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to th9 integrity of a fission product barrier. _ X (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to put'lic health and safety. If the answer to any one of the above questions is "YES", then the T3 shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. Equipment required to perform this function includes, depending on operating conditions, a combination of: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power source from operabh diesel generators. The boration subsystem of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to SDM. To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the RCS, and J appropriate charging pumps to provide the necessary charging head. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA.or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel l7 design limits will not be exceeded for normal shutdown and anticipated operational Attachment 21 9 _______________D

occurr:nces. Th3 SDM d fin:s the degree of subcriticality th t would b3 obtaintd immediately following th] ins:rtion or scr:m of til shutdown and control rods, assuming that the single rod assembly of highest worth le F 'y withdrawn. When the unit is in the shutdown and refueling modes, the SDM requ! sents are met by means of adjustments to the RCS boron concentration. The boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1. The bmation subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem TS does not satisfy criterion 2. For these events, the primary success path for mitigetion includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. The boration system function at shutdown is used to provide reactivity control, in particular for dilution events. As part of the CPSES Safety Monitor shutdown model development, the issue of dilution events as an initiator to boiling or core damage was investigated. This investigation concluded that dilution events are essentially insignificant to boiling or core damage, and therefore, these events were not included in the model. This is consistent with the results of other plant shutdown models, including the Cutage Risk Assessment and Management (ORAM) model. Therefore, reactivity control, and thus the boration system function, is concluded to be an insignificant contributor to shutdown risk for CPSES. Given the foregoing, it is concluded that the boration flow path SSCs are shown NOT to be significant to public health and safety. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION This Technical Specification is retained.

               .X.       The Technical Specification may be relocated to a licensee controlled document.

l l l i t l Attachment 21 10 j t i

TECHNICAL SPECIFICA flON SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.2.6 BORATED WATEP SOURCES - OPERATING Applicable MODES: MODES 1,2,3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant I pressure boundary. l _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis i that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ X (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled doctment. (3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each mode of facility operation. The equipment required to perform this function includes, depending upon operating conditions, combinations of: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from operable diesel generators. The boration subsystem of the CVCS provides the means to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the RCS and to help control the boron concentration to maintain SDM. To accomplish this functional requirement, the boration systems TS require a source of borated wcter, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head. The boration subsystem is not assumed to operate to mitigate the consequences of a , DBA or transient. In the case of a malfunction of the CVCS, which causes a boron l dilution event; the response, or that required by the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Operation of the boration subsystem is not assumed to mitigate this event. Furthermore, Ref. 3 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. f SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational I Attachment 21 11

occurrenc s. Ths SDM d fin:s th3 d: gree of subcriticality th t would b3 obtained immedittily following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. Based on the foregoing, the boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1. The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Therefore, the boration subsystem is not a design feature required to be operable to mitigate these events, and this TS does not satisfy criterion 2. For these events, the primary success path for mitigation includes isolating the dilution flowpath. The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. This is desirable, but beyond the scope of a primary success path action. The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST and associated flowpaths is required as part of the ECCS TS. For the MSLB event, the sequence of events takes the plant to cold shutdown conditions and; therefore, boration of the RCS is necessary. However, the boration flowpath in this case is required as part of the ECCS function. The boration function of CVCS at operation is explicitly modeled in the CPSES Individual Plant Examination (IPE). The emergency boration function is modeled specifically for mitigation of the anticipated transient without scram (ATWS). The IPE model of this function includes both the boration path from the boric acid transfer tanks, via the boric acid transfer pumps and the path from the refueling water storage tank via the centrifugal charging pumps. Failure of the emergency boration function for ATWS mitigation is not a significant contributor to core damage frequency, given the low failure probability and the redundancy and diversity of these flow paths, based on the results of the CPSES IPE. In addition, the ECCS function of the Refueling Water Storage Tank is explicitly modeled in the CPSES IPE and is being retained in the CPSES ECCS Tech Spec. Because the ECCS function of injection from the refueling water storage tank via the CCPs, and l therefore one of the emergency boration flow paths, remains part of the improved tech l spec, the emergency boration function is still assured. Thus, it can be concluded that this tech spec does not satisfy criterion 4. (4) CONCLUSION _ This Technical Specification is retained.

2. The Technical Specification may be relocated to a licensee controlled document.

Attachment 21 12

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.1.3.3 POSITION INDICATION SYSTEM - SHUTDOWN Applicable MODES * **: 3,4 AND 5 (* With the reactor trip breakers in the closed position. See Special Test Exceptions Specification 3.10.5.) (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YES NO _ 1 (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary, _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. l If the answer to all four of the above questions is "NO", the TS may be relocated te a controlled document. (3) DISCUSSION l The Digital Rod Position Indication System verifies the agreement with demanded ! position at specific steps within the rod position range. ! The Digital Rod Position Indication System is not an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Digital Rod Position Indication System TS is not applicable to installed instrumentation tnat is used to detect and indicate in the control room a significant abnormal degradation of the RCPB. Therefore, this TS does not satisfy criterion 1. The position indication system TS, for shutdown conditions, is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challeng6 to the integrity of a fission product barrier. Therefore, the TS does not satisfy criterion 2.

 .                  The Digital Rod Position Indication System TS does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or Attachment 21 13

transiint that either essum:s th3 failurs of or pres nts a chill:nga to tha intigrity of a fission product birri:r. Thira fore, ths TS does not satisfy criterion 3. From Ref. 3, the Digital Rod Position Indication System has not been shown to be a significant risk contributor to public health and safety by either operational experience or PSA. The Digital Rod Position Indication System is not modeled in the CPSES IPE. This TS does not satisfy criterion 4. (4) CONCLUSION _ This Technical Specification is retained.

                                                                                .2L      The Technical Specification may be relocated to a licensee controlled document.

l Attachment 21 14

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP-3.1-ED APPLICABILITY: CP REQUEST: Minor corrections / clarifications / editorial changes / consistency changes ATTACHED PAGES: EnclSB B 3.1-4 Ref. 3 changed to Ref. 2 B 3.1-28 Added words "of Condition B" for clarity B 3.1-38 Ref. 5 changed to Ref. 3

                                                                                                            'l J

SDM -T,,, r 200 i B 3.1.1 BASES terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits. ThestartupofaninactiveRCPjniModes;11and[2jjs;precludedfby admi_nistrative; procedures jIn1ModeMthelstartup_ofianlinactive RCP"can;not result in a " cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition. The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of core power. SDM satisfies Criterion 2 of the NRC Policy Statement 10CFR50;3.6(c)(2)(11). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions. LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

                                                                      ~

The MS ,f,. m 2) and the boron dilution (Ref. 3) accidents m CP 3.1 ED (Ref :are the most limiting analyses that establish the SDM v ue of the LC0. For MSLB accidents, if the LC0 is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LC0 is violated, the minimum l required time assumed for operator action to terminate dilution j may no longer be applicable available.[:Jhefrequired;SDMiis specifiedlin'the COLR; (continued) CPSES Afark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-4 7/29/98

Rod Group Alignment Linits B 3-1-5 3;174 BASES 1 ACTIONS determine that core limits will not be exceeded during a Design I (continued) Basis Event for the duration of operation under these conditions. The accidentianalysesl presented;in FSAR; Chapter 15;(Refn3)Rthatimay;beradver_sely;affected will;be, evaluated 0 3.1 13 toiensure;thatLthe analys.esiresults; remain yalidiforithe j durationloficontinuedloperation,under;these; conditional l:A I Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis. B-1-C :1

                                                                 #                         T WhenRequiredActions(Condition            annot be completed        CP-3.1 ED
                                                                                                         )

within their Completion Time, the unit must be brought to a  ! MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at { 1 least MODE 3 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from i'ull power conditions in an , orderly manner and without challenging the plant systems. j C.I.1 cr:d 0.1.2 D:1?1#and 0.1:2 More than one control rod becoming misaligned from its group cycrage demand? position is not expected, and has the potential to reduce SDH. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LC0 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the beric acid required pumps. Boration will continue until the required SDM is restored. C-4-0. ? If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. (continued) CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-28 7/19/97

Control Bank Insertion Limits B 3-1-7 3.1;6 BASES APPLICABLE flow, ejected rod, or other accident requiring termination by an 4 SAFETY ANALYSES RTS trip function. (continued) The acceptance criteria for addressing shutdown and control bank insertion limits ad inoperability or misalignment are that:

a. There be no violations of:

1. specified acceptable fuel design limits, or

2. Reactor Coolant System pressure boundary integrity; and

b. The core remains subcritical after accident transients.

As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3). The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivity is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref.4 3). Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed OPTR present. Oparation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths. The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and pow distribution peaking factors are preserved l CP 3.1 ED (Ref. 3) l Implicit;1niallical.culations which involve;the:banklinsertion limits).isitheiassumpti.onithatinormal; control; bank 3equence and overlapfare_maintainedi The insertion limits satisfy Criterion 2 of the NRC Policy State: nt 10CFR50.36(c)(2)(11), in that they are initial conditions assumed in the safety analysis. L (continued) CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-38 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Gr)-3.1-002 APPblCABILITY: CP REQUEST: Revised traveler page to reflect latest revision numbers and/or approval status. l l ATTACHED PAGES:

                                                                                                                                           )

Enci5A traveler pages (2) 1 I l

Industry Travelers Applicable to Section 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF 9, Incorporated 3.1-1 NRC approved. Revision 1 _ M 10 Het ir,corporated NA NP,0 rcjccted. ( P.3.1 002 [n _ . ,m. 2. 2. u_,_,. t TSTF-11, Not ir,ccrporated -NA N",C rejected. ( P 3.1 002 A

                                  \                          n_.2_2__
                                                                  . s .a i vi s A TSTF 12,               Incorporated           3.1 15          NRC approved. ITS Special Revision 1                                                      Test Exception 3.1.10 is retained and re-numbered as 3.1.8 consistent with this                     i traveler and TSTF-136.

TSTF 13 Incorporated 3.14 NRC approved. , Revision 1 TSTF 14 Incorporated 3.1 13 NRC approved.S ith tiOCE c p.3,1 002 l Revision --34 s't4e+ned ir. ITS 3.1.0 applicability. M 1 R 3.1-005 TSTF 15, Incorporated NA kapproveddifct yet apprs' , ( p.3,1,002 Revision 1 '^gacI1cally. TSTF 89 Incorporated 3.1 8 NRC approved. TSTF-107 Incorporated 3.1-6 .mC requested char.ges3 , p.3,3 002 M TSTF-108 Not incorporated NA Net NRC approved. es-ef ,g.3,1 001 Revision _1 3.1 21 traveler cutoff date-TSTF 110. Incorporated 3.1 10 NRC approved! , g.3.1.co4 Revision'2;f Kic s.... ..,. Navel e' justificaticr, per N",0 ( gr,ts. ( p.3.1 002 TSTF-136 Incorporated 3.1 ..;3.1 15 _

                                                                                                                                                           , c p.3,1 002 l                                                               TSTF 141             Not incorporated           NA            Disagree with change; traveler issued after cut-off date.

t

TSTF-142 Net-Incorporated 371E22.NA NRC~ approved OTeavelcr issuaw 1 R 3.1 003 eftee-eut off date-

                                                                                                                                                                         -- ~

WOC 73; g;:/. ,

                                                                                            ..                    _ Incorporated                                   3.1-7                                                       -

( P-3.1 002 ISTE-  ! 1R 3.1 006 l W0G 105 Incorporated 3.1 16 ) I 1 l

i ADDITIONAL. INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: CP-3.1-003 APPLICABILITY: CP REQUEST: The phrase "at hot zero power conditions" was added in error during the original markup of this Bases section. It is now being deleted.  ! f(TACHED PAGES: Enci5B B 3.1-3 i i i 1 l i l I i

i SDM --T,,, r 200 I I B 3.1.1 l BASES I l APPLICABLE coolant shrinkage causes a reduction in pressure. In the i SAFETY ANALYSES presence of a negative moderator temperature coefficient, this (continued) cooldown causes an increase in core reactivity. As theLinitial RCS temperature decreases, the severity of an MSLB decreases until the-MODE 5 vake-is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs,isaguillotinebreakofamainsteamlineinsig CP 3.1 003 contai mont initiated at the end of core lif atf.ctjep .

                                             . Ine positive reactivity a                                 6 ion fi um temperature decrease will terminate when the

{' affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1. In addition to the limiting tiSLC transient, th: 50t1 rcquirc; cat l aust also protect against; {

c. Inadvcitent borca dilution,

b. An uncontrolled rod withdrawal fic; subcritical or icw i pcwcr condition.

c. Startup of an inactive reactor coslant pump (RCN , and

d. Rodcjection.-

Each of thesc cvcats is discussed below. In the boron dilution analysis, the required SDM defines the reactivity difference between an initial suberitical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.L The shutdownmarginmustbe; adequate;toallow; sufficient; time;forithe r_eactorE. operators;to; detect;an 1_nadvertent boronJdilution~ and initiateLappropriatel action;to;provent a; complete;10ssiof l shutdownimargin] l Depending on the system 1iltial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is i l 1 ( (continued) l l CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-3 7/29/98 l l

l ADDITIONAL INFORMATION COVER SHEET I ADDITIONAL INFORMATION NO: DC-3.1-001 APPLICABILITY: CP, DC REQUEST: Revised DOC 12-11-TR-3 in enclosure 3A to reflect lead plant wording i l l ATTACHED PAGES: Enci 3A 8 t 1

                                                                                                                                                                                                                                                     ~

i 1 l i l f

CHANGE NUMBER EliC DESCRIPTION accidents to assure the results remain valid. This is acceptable because the general requirement is sufficient ( to assure that the affected accident analyses will be l considered and the ITS Bases discuss the accident analyses affected by rod misalignment. l l 12-08 LS-9 Consistent with NUREG-1431, Rev. 1, the requirement to reduce the high neutron flux setpoint to s 85% of RATED THERMAL POWER (RTP) would be deleted. This is acceptable because the underlying safety limits are not of a nature l that requires immediate shutdown of the plant if they are exceeded. This is evidenced by the allowance of 72 hours to verify peaking factors. It is assumed that during this ! 72 hour period an event will not occur which will raise the power level and cause a high neutron flux trip at 100% RTP. If a power excursion would occur from the 75% RTP Action Statement limit, the initial peaking factors would not be critical to the analysis since the analysis is based on the peaking factors at 100% RTP. Therefore, the risk of a reactor trip caused by adjusting the power range trip setpoints is not justified by the potential consequences of failing to reduce the trip setpoints. 12 09 M Not applicable to CPSES. See Conversion Comparison Table (Enclosure 3B). l l 12 10 LS 10 The requirement to maintain RCS T,,2551*F during rod drop testing would be revised to maintain T.,,2500*F. NUREG-1431. Rev. 1 allows the tests to be performed at temperatures as low as 500*F. Because the RCS coolant is more dense at lower temperatures, the rod drop time would be greater at the lower temperatures than at the higher temperatures. In addition, the RCS is borated such that I the SDM remains within its limits at the conditions ) existing during these tests. Nevertheless, this change, which allows more flexibility on plant conditions for conducting rod drop testing, is a relaxation in plant operations under tne TS. 12 11 TR 3 It is proposed to remove from the current TS the requirement to perform rod drop testing on individual oc.3.1 F rods following maintenance that could affect the drop time to licensec contrclicd documcats [and to delete the 18 month requirement]. The requirement to perform drop time testing following each removal of the reactor vessel head would not be modified. The proposed change is I justified because, in addition to being consistent with NUREG-1431. Rev.1, good maintenance practices would require a retest l CPSES Description of Changes - CTS 3N.] 8 7/29/98 l

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR-3.1-001 APPLICABILITY: CA, CP, DC, WC REQUEST: Incorporate NRC-approved traveler TSTF-108 Revision 1 to delete the words "within 12 hours" from the Frequency of CTS SR 4.10.3.2 and ITS SR 3.1.8.1. ATTACHED PAGES: Attachment 16 - CTS 3/4.10 Encl 2 3/4.10-3 EncI 3A 2 Enci 3B 2 Encl 4 1,15 & 16 Attachment 7 - ITS 3.1 Enct5A traveler page and 3.1-22 Enc!SB B3.1-55 and 56 Encl 6A 4 Encl 68 3 l l L_________________________ _.

SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITED CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3. 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The ik;ctor Trip Sctpcirts en the OPEPASLE Intcrecdiate and Powcr i 3-02-A l'sagc chonnels arc set in accordance with Tabic 2.21 Punctional Units 5 and 2b. and

c. The Reactor Coolant System lowest operating loop temperature (T.,,) is greater than or equal to 541*F[and pew The~ SHUTDOWN MARGINfisLwithin;thellimitsLspecifiedl.inithe COLR. L3-01-M a APPLICABILITY: MODE 2 during PHYSICS TESTS. [3-04-Aa  ;

ACTION: l

a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers,

b. With a Reactor Coolant System operating loop temperature (T,,,) less than 541*F. restore T.,, to within its limit within 15 minutes or be in at least H0T STANDBY within the next 15 minutes. I new With the: SHUTDOWN l MARGIN lnot Within its 11mitsl1within_15 minutes; i 3-01-M -

initiate:boration,to; restore the; SHUTDOWN MARGIN to:within;itsLlimits and ,within 11hourisuspend fHYSICS TESTS exceptions; SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of l RATED THERMAL POWER at least once per hour during PHYSIrS TESTS. l 4.10.3.2 Each Intermediate and Power D ca channn1 shall be subjected to an 3 05 LS ANALOG CHANNEL OPERATIONAL TES .athin 12 hours prior to initiating PHYSICS u 3.1 001 l - TESTS. 1 l I 4.10.3.3 The Reactor Coolant System temperature (T,,,) shall be determined to be j greater than or equal to 541'F at least once per 30 minutes during PHYSICL TESTS. pewi. T !. Verify the. SHUTDOWN. MARGIN:1s;within the 1imitg specified in the COLR I every;24 hour _s during' PHYSICS;TfSTS; 03-01-M CPSESMarkup ofCTS 3N.10 3M.10-3 7/19/98

l 3 CHANGE NUMBER NSHC DESCRIPTION l change is acceptable because there is no need to reference another LCO if that LC0 is applicable in the same MODE. t 3 03 Not used. l-3 04 A The applicability statement would be changed to be more consistent with operation for testing purposes. The proposed change is consistent with NUREG 1431. Rev 1 and does not result in any changes to technical requirements. l 3105 LE1 The; current 1SRJrequirjagit heZperforsangeloft[an n 3.1 001

                                       $NALOG):CHANNELLOPERATIONALETESTionzeacNjntermediate

( and: power 1rangeiNISichannet;within1121 hours:prjorito '

                                       ,1nitiating fHYSICS;IESTs11 sireyiseditoldelete;$heArase twith1E 12; hours;5 Current 1TSlLCO:373 E1;iReactor11tip System 3RTSEinstrumentation;1 requires;the: performance;of l                                       ani[A]COTEogithe; power; range 3ow;setpojpt[andiintermedjate l

tange:NIS; channels;prjor;toleachiteactot;startupHjfinot performed;withjntthelpteyiousr31daysi(revised;to;92; days ]

                                       .inithe; conversion 2to;ITS13;3)RThese'RTS1SRs;sust;be                    J l                                       petformedf ptioratolenteringithe:LC0;3;hli Appljcab1]ities                ;

i forJhesetRTSitrjpfunctions1si.nce;thete:arelpoiCTS1Sg l 4:;0Leexceptions Ecorrent[SR W10@2;teqtriteLan l arbitrary; estimate;of;whenj.thel plant M s;With k 12; hour Cof j nitj atjng; PHYSICS 3ESTING EThi sihas! no; basi sifron 'the 4 accidentanalysestiwhich;are;;satisfiedjas2)onglasithe sutveillances;arelopreentl;ptiotitoienteting plant; MODES  ! where these; trip; functions ; provide: protection E Whenithese suryeill ances;areicurrent;0 theyihavelpreviously, been determined i toiremain;yalid:fot;92;daysChel.initiationiof RHYSICS? TESTING (doesinot ;1mpacts the rab111tyLofithe channel.s;to"performithejrrequiredfunction;idoes;;not l affectithe;tripisetpoints;ottr,1pi. capability of;these i channels;1and;doesinot) invalidate the previous surveill ances EThi s; change Li s';consi stent;withitraveler TSTfd108I l 4-01 M Special Test Exception [LC0 3.10.4] would be deleted. This specification allows the suspending of requirements of one or more LCOs (depending on plant specific current TS) under certain conditions. Elimination of the special test exceptions is justified either because their elimination would be consistent with NUREG 1431 Rev.1, or because the applicable tests are performed only during initial plant startup and are no longer needed. 1 CPSES Description of Changes to CTS 3N.10 2 7/19/98 E__________________ _ _ _ - - -

a W - ee ee P A rm r m L d d L - n - n s s e e A e e om om C Y Y NA NA r r e p e p d d e e K t 9 t 9 E a8 a8 c c E ot ot Y R l n l n C ee ee T rm rm I F d d L L - n - n s s e e I O e e om om B W Y Y NA NA A C I L M - 0 P 2 1 P K 0 A A 5 4

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ml i uo prt t i ,o. i t p w ti co i r r 'e cb i t p wirr t coocl i n E eNl c o enpa p e e T r,I Onies xd E es n t t oat ea scr c",sSoThd xnP c".sS oTh o t n ExnP i tl RT n t t el t e xa E w w f i - N SAaS RhT t enra ah t e sl eri nt n ] td o ", foiwu oc td o ", o w s i O tE tS eemuci a st3 nd st 3 nd e 4. d I nP E Td ecemel r h eu oe5 eu oe5 T eOST pi s pt0 e Th3.id P r r L' I l eicsl u 1 t a S t u0 Th3.idS t u0 I R uE NS C T. i ab ui qf e eal e ca s3 n l are 1 al c c 1 l a m 1 al cc 1 C cI eI s cd eih r ei u i i e3i n3 i e3 i n3 MgS r el r ctibt aO m ct l i ct l i S eA nY u pu e e i cCi esO p O esO p O E hH aH o S of pfh r neLl pyC peC pyC p eC D TC rPh wos ot oib[ e SSL abL SSL abL R E B 7 51 _ 1 1 2 _ M 0 - 0 0 0 U - S - - - _ N 3L 4M 5R 5M

t i-I l. N0 SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) CONTENTS l l I. O rga n i z a t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 II. Description of NSHC Evaluations........................................ 3 l III. Generic No Significant Hazards Consideration l

                            "A" - Admi ni strati ve Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 "R" - Relocated Technical Specifications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l                            "LG"       Less Restrictive (Moving Information Out of the Techni cal Speci fi cations) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 "M"       More Restrictive Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 TR-3.1 001 IV. Specific No Significant Hazards Consideration                                               "LS" NoneLS ':.1.*.s :. r".. ru~a~n"? . . , 'v". . ."."."u'. .*^              .m" ^...        . e"s . . '? w' t'. . '. n"u" .^. . ~, '. .. . '.15 l

[ I l l i l CPSESNo SignifIcant He:ards Considerations - CTS 3M.10 1 7/29/98 1

E _. " IV. S,PECIFIC1N0;SIGNIFICANTLHAZARDS CONSIDERATIONS LA T' .'s '

                                         ,      _1       NSHCli.S!1 U_,'                               S10 ECFR250;92 EVALUATIONS

_ '._ Z ,, ,;FOR

                       ~

I . '_ .. _ r ;TECHNICALTCHANGES1THATJIMPOSE_LESS, RESTRICTIVE

         --           m       :
                            ' REQUIREMENTSJWITHIN1THE1        TECHNICAL > SPECIFICATIONS JheicurrentiSR requiring:the performance:of [an1 ANALOG]:CHANNELLOPERATIONAL                                                m.3.1. col TEST on'each~1 intermediate;and; power., range;.NIS; channel.1within 12 hours prior                                 n to initiating PHYSICS :TESTSt is; revised 1to delete <the' phrasei"withi.n; 12: hours' ." :.This                     .

change;1s_consi_ stent with traveler lTSTF)108 This.. proposed TS; change has;been: evaluated:;and Litihas been ' determined. that'it; involves no;;.significant; hazards [ consideration.7Thisfdetermination ha.s been. performed;in accordan.ce.with tt;e criteria (setiforthlin:10;ECERf 50.92(c);;as quoted below;

L , "The Cormdssion"maymake:a[ final deterninationVpursuantLto the procedures;1n, ;

50. 91 V tha t aJ proppsed;amendrnent to ' an opera ting license Lfor:a Jfaci l ity J11 censed / under;50.21(tt) or" 50;22;prl fo(a[ testing: faci li ty L1nvolves; no sign 1ficant hazards: consideration; if Operation of the; facility in;accordance With the; proposed amendmentLwould not;b I

1; Involve a significantiincrease;in'the probability;orjconsequences,of;an s acciden t; pre vious ly;eva lua ted; Lor 2: Create the possibility pf;a Aew;or:different kind /of' accident fran any i accident ' previous ly;eva lua tedh on 3: :1nvolve a significant reduction:1n;a margin of safety.", a The;following evaluation ,isiprovided for;the three; categories,of, the..significant hazards considerationrstandardst

1. Doesithe change involve a significant increase-in the probability or consequences.of.;an; ; accident previously _ evaluated?1 ~

Overall . protection; system performance:will remain.within;the. bounds of;the previously performed accidentLanalyses;since.no hardwar.e' changes are proposed.' Current TS.LCO 3.3.1L Reactor; Trip System (RTS) Instrumentation,; requires the performanceLofian-[A]C0T;on the power; range low:setpoint and intermediate range , NISTchannels prior _ tojeach ' reactor startup,11f'not performed.within.the l previous;31 days;;(revised;to:92: days;.in,.the conversion to ITS 3.3) CThese RTS ) SRs must;be performed prior,to: entering the LC0 3;3.1 Applicabilities for, these l RTS: trip functionsisince there are no1CTSiSR14.0._4 exceptions. Current SR . 4.10.3.2Lrequires' an arbitrary; estimate of when'the plant:is;within 12 hours of i initiating PHYSICS; TESTING This;hasino basis from the accident analyses. which i are satisfied as long.as the surveillance are current prior to entering plant H0 DES;where;theseLtrip functions provide protection.iWhen these surveillance l CPSES No Significant fla:ards Considerations - CTS 3N.10 1S 72988

EE& HIVMSPECIFICIND;SIGNIFICANT, HAZARDS _ CONSIDERATIONS n y ,, n-

                                                                                  ,n g ,3 ,g eg;,

g $bdbNdh kkN[ continued) are ' current G theys have; previously_ been;; determined;to ;remainival id;for; 92

                                                                .                                                                                  Tg.3,i.coi days tThe ,initiationiofLPHYSICS JESTING;doesi notiimpact ithelabil ity;of the; channel s' to ;' perform;thejr, requiredffunctionildoes:not; affectithe ' trip setpoints oritripleapability ofithese channels;iand[does;notlinvalidate<the previous surveillancesH Thel proposed cha_ngelwillLnot: affect;any;ofsthe; analysis assumptions :;forzany;off.t.he;; accident _s;prevjously evaluatedGThe: proposed (change

, wil1Lnot affectithe; probability;offany3venttjnitiatorsinotiwilhthe: proposed  ; ! change." affect 1the ab111tyLoffany; safety related:equipmentito;perforelits intended; function 3There;w111[be:no; degradation;in;the:performancelotnordan 19ereaset.in the; number;ofzcha11enges;imposedion;safetytrelated equipment as.sumed to function;during:an'accidentEsituation EThereforen thel proposed change doesinot; involve;alsignif.icantlincreaselin;the; probability or consequences ofZan; accident previously;evaluatedi 1 2: Does the: change createithe-possibility;of;alnew ors different: kind lof ) accident;from any;accidentSpreviously; evaluated? i There are,no hardware;changesinoriareitherelany; changes;in;the method by which l any; safety related;plantisystem; performssits;safetyifunctionOThe1 change;toLthe i SR1w111:not; affect lthe; norma 11 method"of;plantioperationnNo new accident . scenarios ; transient; precursors;;fa11ure mechanisms,;or: limiting ' single  ! l failures;are introduced las;airesultiof:this:changeRTherefore,Mthe; proposed ! change"doesLnot createithe7 possibility;ofcainew ot differentikindioffaccident from;any previously: evaluated; j 3; Does.;this changelinvolve;arsignificant red,uctionz infa margin offsafety? l l l The proposed changeLdoeslnotEaffect;the acceptance: criteria r analysis assumptions ' ; methodologies Mor; credited ;equipmentifor;any; analyzed; event CThere Will belno..effect..on1the manner 31n which:safetyi. limits;or a linitingjsafety system; settings;are: determined;nor;w11Ethere.,belanyfeffect on:those plant systems;necessary toiassure;thelaccomplishmentlof@tection1 functions %There w11Ebe no impact: on;any marginiofesafety; E Z M N0;SIGNIFICANT;HAZAP.DS CONSIDERATION; DETERMINATION :7 Based on the;above evaluation Rit;isiconcluded that::,the activities 1 associated with NSHC/LSiltresultinglfrom;the; conversion lto the.risproved:TSiformat; satisfy theino i significant hazards; consideration: standards lof(10ZCFRl:50.,92.(c);;and.accordinglyCa nonsignificant hazards; consideration (findingsisijustified; I CPSESNo Significant Ha:ards Considerations - CTS 3M.10 16 7/29/98

                              -- - - - - - - - _ . . . - - - - - - - , - - - - - - - - - - - - - - - - - - - , - - - - - - - - - - , - - - - - - - - - - - - ' - - ' - - ---'--                    ' - - - - - - - - ~ ' ' ' - " - - - - - - " - - ' - ~ - - " " ' ~ - - ' - ~ ~ ' ' ' ' - ~ - - - - - - --                  --

Industry Travelers Applicable to Section 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF 9, Incorporated 3.1 1 NRC approved. Revision 1 TSff Not incorporated NA NRC rejected. (P-3.1 002 Revision-4 1 TSTF 11, Not incorporated -NA NRC rcjccted. (P 3.1 002 Revision i TSTF 12 Incorporated 3.1-15 NRC approved. ITS Special Revision 1 Test Exception 3.1.10 is retained and re-numbered as 3.1.8, consistent with this traveler and TSTF 136. TSTF 13. Incorporated 3.14 NRC approved.  ! Revision 1 I l TSTF 14, Incorporated 3.1 13 NRCoapproved. with liOCE 2 (p.3,1 002 l Revision --34 retaincd in !TS 3.1.8 A - 1 J A J 1 4 4. . . DFV' "uw 'V-1R 3.1 005 TSTF-15. Incorporated NA NRCLapproved.: Not yct approve jcp.3,3 002 Revision 1 gcacrically. =-.- { Incorporated 3.1 8 TSTF 89 NRC approved. TSTF 107 Incorporated 3.1 6 NRC requested changes. cp.3,1 002 l l

                             -                                                                                                                                                                                                                                                                                                      1
                     ,/                                                                                                                                                                                                                                                                                                   _A
                                                                                                                                                                                                                                                                                                                                    \

Net Incorporated Net NRC approved. es-ef TSTF-108 Revisionf1 NA 3,1-21 travelcr cutoff date. h R 3.1 001 l 1

                                                                                                                                                                                                                                                                                                                         ]           1 TSTF 110                                                                                                    Incorporated                                  3.1 10    NRC'approvedi                                                                                                       3g.3.1 004 Revision'2 3                                                                                                                                                            TSTF to enhance travelcr justification per NRC                                                                                               ( p.3.1 002 cc.. cats.

l TSTF 136 Incorporated 3.1 9, 3.1 15 (p.3,i.002 TSTF 141 Not incorporated NA Disagree with change; traveler i issued after cut off date. TSTF 142 Not--Incorporated 3.1-22 NA NRC; approved. Travcicr issucu 73.1003 l cfter cut-off date.

                                                                                                                                                                                                                                                                                                              --            _a

PHYSICS TESTS Exceptions-MODE 2 3 .1 - 3.1.10 3.1;8 3.1 ACTIONS (continued) CGNDITION REQUIRED ACTION COMPLETION TIME C. RCS lowest operating C.1 Restore RCS lowest 15 minutes 3.1-20 loop average temperature operating;1oop average 0 3.1-17 not within limit, temperature to within limit. D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.10.13.1.8.1 Perform a CHANNEL OPERATIONAL TEST on W4tMti y3],gg1 power range and intermediate range W-k l channels per SR 3)311.7;, SR[3.3(l'.6;'.and gor to , Table' 3.311F1. Titiation B of PHYSICS TESTS SR 3.1.10.2 3;1.8.2 Verify the RCS lowest operatingcloop 30 minutes average temperature is 2-E5343 541*F. 3, l'f17 B-PS t 1 sri 3.L1'.8.3 Verify:THERMAll POWER is i 5% RTP. I hour SR 3.1.10.3 Verify SDM is < 1.FA ok/k. Within theLlimits 24 hours 3.1-1 '

     ' 13.1'8._41        providedlin_the.COLR; CPSESMark-up ofNUREG-1431 - ITS3.1                      3.1-22                                                                                                               7/29/98

I PHYSICS TESTS Exceptions-MODE 2 l B 3.1.106 l BASES Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification. IL1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits. fa.1 When the RCS lowest [operatingil.oop T,,, is < 531 541 F, the 0 3'1-26 appropriate action is to restore T.,, to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring T.,, to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 531 541*F could violate the assumptions for accidents analyzed in the safety analyses. D.l If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems. , SURVEILLANCE SR 3.1.810.1 REQUIREMENTS l The power range and intermediate range neutron detectors must be l verified to be OPERABLE in MODE 2 by LCO 3.3.1, " Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performad on 91sch power range and intermediate range channel, TR 3.1 001 Cethir,12 hougprior to initiation of the PHYsich TESTS. _ inis win ensure that the RTS is properly aligned to e the required degree of core orotection during t erformance of the PHYSICS TESTS. hour ti c limit (continued) CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-55 7M9/98

PHYSICS TESTS Exceptions-MODE 2 B 3.1.M8 BASES l is sufficient to casure thabthe instrumentation is OPEPM l -hcrtly beforc initicting PllYSICS TESTS. . n.3,1.coi SR 3.1.M8.2 l Verification that the RCS lowest operatingjloop T., is 2 531 l 0 3.1 26 541"F will ensure that the unit is not operating in a condition that could invalidate the safety analyses. Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the l initial conditions of the safety analyses are not violated. I SR'3'.1.8.3 Veri fication ' thatitheJTHERMALEPOWER ^1 s9 StiRTP, will : ensure, that;the plantLislnot; operating'infascondition1thatTcould invalidate;the: safety analysesGVerification ofl:the2THERMALLPOWER"atla frequency ofl11 hour duringLthejperformance;off.the1. PHYSICS.; TESTS'wi1Eensureithat the l initial! conditions;of;.t.he; safety analyses;are not' violated. SR 3.1.M8.43 Verifi. cation that the;SDHiisiwithin lini.ts;specified in theLCOLR ensuresithatiforlthe; specific RCCA'land .RCS;.temperatureLaanipulations , performed [duringLPHYSICS;TESTSEthe plant;isinot; operating jn .a l conditionithatlcould; invalidate the L. safety;; analysis assumptions;.JThe l SDH verification can be facilitated throughf the:Uselofitables~ prepared f by:thelcoreldesigners:inlwhich:thefreactivity;.effectsTexpected1during thel. Physics; Testing have been;previouslymconsidered; The SDM is verified by performing a reactivity balance calculation. I considering the following reactivity effects:

a. RCS boron concentration;

b. Shutdown..and Control bank position:

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation:

! e. Xenon concentration: l

f. Samarium concentration; and l

g. Isothermal temperature coefficient (ITC). '

l l l CPSESMark-up ofNUREG-1431 Bases - ITS3.1 B 3.1-56 7/29/98 i i l j

l l CHANGE NUMBER JUSTIFICATION 3.1 18 A MODE change restriction has been added to ITS 3.1.1. in the LC0 applicability, per the matrix discussed in CN 1-02 LS-1 of the 3.0 package.  ; I 3.1 19 Not Used. I 3.1-20 Consistent with current TS 3/4.10.3, " Physics Tests," ITS LC0 3.1.8 and its Condition C and SR 3.1.8.2 are modified to refer to " operating" RCS loops. Adopting the current TS is i acceptable since valid Tavg measurements are not obtainable for a l non operating loop. -_-__ I 3.1921 The ISTS[,SR 3;1(.8:1;:' requirement tofperform;a

                                                                                                                                                                                                                                                                         .                c             CHANNEL              TR 3.1-001 ) {

OPERATIONAL 1TESTd(C0T)1on the; intermediate l'and power. range

                                                                                                                                                                                                                                                                                                                                          )

l NIS channels;Within 121 hours priors to jnitiating, PHYSICS  ! TESTS il s Lrevi sed l tof del ete;. the : phrase;"withi n 12 ; hours . " . COT testing;1s; performed orr.;these.; channels; prior to.; reactor; startup per LC0 3;3.^1621s; change:js consistent;with TSTF 1 __, j 3.1 22 TheLCompletionfTimes;for ITS.3.1.2,LRequired_ Actions A;1(and 7g.3,1 003 A.2;arel increased from:72 hours;tol7fdays,; consistent with traveler TSTF 142l I 1 I 1 l l  ; i i CPSESDifferencesfrom NUREG-1431 -ITS3.1 4 7/29/98 [ E_________________________ _ _ _ _

C Y N Y Y y K E E R C Y F _ T L s A s s _ _ s I O e / e e _ e L W Y N Y y y I B - 1 A C K I A 3 L E N P P O P I A E H T C C N E A S M s A s s s O e / e e e

     ,   C Y          N     Y            Y                           Y 1

3 4 1 N O G Y E N R A U C N _ O M L O B R A s A s s s F I e / e y e e D Y N Y Y S E C e N g ts . ne r E a i s i p p R k c d o Tst p h? E a n o F p a l S E u riu e F I 0 8 S C T o "w rat h r a O L D C3 L 1 R A2s N1 set 2" R O e 3 "g O : ra: o A eh ht O n I Ti nh t d' F t C i Ah p ;r n L ta Rt a Eih e o i f E no S r r L i T ep P wt r I B ,1 0;se . Jp A - I A T 1 S L ". t Ll E e)l t s s 1 - s o N r e n N 2 t t N n n o O S 30 S1 s e T ef r A Hh a@m C. c to ch i t c I T A R I N C s c e r aSl

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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR-3.1-003 APPLICABILITY: CA, CP, DC, WC REQUEST: Incorporate NRC-approved traveler TSTF-142 to increase the [] ITS 3.1.2 Required Actions A.1 and A.2 Completion Time from 72 hours to 7 days when the core reactivity balance is not within its limit. 1 ATTACHED PAGES- l EncI2 3/4 1-2 1 EncI3A 6 Enci 4

                                                                                                                                                                                                              ]

48 and 49 ' EncI SA Traveler page and 3.1-2 En .I 5B B 3.1-12 Enc.SA 4 l Enci 6B 3 l

                                                                                                                                                                                                              ;1 1

l l l i I i 1 1 i 1 , \,

 -                            _-_- - -                                                                            - - - - - - - - - -       -     --                               --                         ]

i 1 REACTIVITY CONTROLS SURVEILLANCE REQUIREMENTS (Continued)

c. When in MODE 2 with K,,, less than 1, within 4 hours prior to achieving 014A reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6:

d. Prior to initial operation above St RATED THERMAL POWER after each .

fuel loading, by consideration of the factors of Specification  ; 054A ? 4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6: and

e. When in MODE 2 with-kg<jl.0, 3 or 4, at least once per 24 hours by E '1-l*

ccasideration of the fcilowing factors.

1) Rcactor Cociant Systc; bcron concentration.  : 'l#W '

2) Control rod positicrn

3) Reactor Cociant System average-tempcraturc,

4) fuel burnup based on gross thermal cacrgy generation.

5) Tenon concentratica, =d

5) S narium concentration.

I 4.1.1.1.2 For Modes 1 and 2 only, the overall core reactivity balance shall be ' 85 8W '

• l compared to predicted values to demonstrate agreement within i li a k/k once prior tofentering Mode 1;after:each; refueling andlat least once per 31 Effective Full 188&IA Power Days (EFPD)!thereafter --This comparison shall consider at least those factors stated in Specif4 cation 4.1.1.1.lc., abovc. Tthe predicted reactivity 05 06 A values sheHeay be adjusted (normalized) to correspond to the actual core 0 3.1-4 conditions. This; adjustment (normalization) sheHeay be rperformed: prior to .

exceeding a fuel burnup of 60 EFPD after_each fupl loading. 'If reactivity . 85 & W balance:is;not within, limits LwithinQ72 ho@, evaluate therSafety - Analyses;and jestabl i shl appropri atef operating estrictionsiand/or surveillance 05 05-ts requirements, Lor:be in at;leastlHode:3 within he nextL6'hoursy TR 3.1-003 l l  ! CPSES Mark-up ofCTS - 3N.1 3M1-2 7/29/98 1 1

CHANGE NUMBER N2iG DESCRIPTION insertion limits is determined through compliance with ITS 3.1.2, which requires a reactivity balance prior to entering Mode 1 after each refueling, and ITS SR 3.1.6.1, which requires a verification of control bank position within insertion limits within 4 hours prior to criticality. Therefore, the requirements of this SR would be performed by other specifications in the ITS. [ ] 05 05 LS 17 Actions to be taken should the reactivity balance not be within the required limits are also provided, in lieu of immediately performing a plant shutdown in accordance with TS 3.0.3. The; proposed: activity wouldiadd m 3.1-oo3 requitediactionsljfithe:overallicoreireactivity balancejas ~ notWithinfilt;Ak/it.:of1the: predicted ValuesM!nithmicurrentJTechi@allSpecificationsRthere areinoirequited;actionstthusGLC0;3i0f3;would1belenteredM LC013.:4.3; requires;that;J withinXhour;1 actions; be initiated 1to:p1_ ace;the;pl. ant 11n a:conditioniiniwhich;the Lcoldid;not; applyE Because;this; partjcular; Surveill ance Requirement Lis: only; required ;in_ Modes 11:and;2,1LC0! 3'. 0 ; 3 l would;furtherl. require 1thatJthej plantibe;placed;in1 hot STAlWBYMMode23);withinithelfo11cwing16;hoursiaThe proposed [changeRconsistent;with 1NUREGi1431LRevisionfij asnoodified:by;TSTE142,1would;a1]ow2 days;to; evaluate the; Safety Analysesland:estab]jshlappropriateloper.ating testrictjensland/otsurveillancelrequirementspIIf;these pctivities were:not completed;within;the;7 day perjod; l thentthelplant, would;be;p] acednin; Mode 13_within1the i follow 1ng'6; hours? Theirequirementito periodical)ylcompareisthe,measuredLand l Predictedovera11Icore; reactivity; balance _s;provides I assurancelthatithefanalytical[predictionslupon. whichithe safetyEanalyses are:basediaccurate;representithe: actual coreitesponseCShould;an~anoma]y; develop;betweenmeasured

                                    .and. predicted;corefreactiyjty gan evaluationiofdthe; core l                                    designiand:safetyfanalysignusttbe: performed;r core l                                    conditions;are:evaluatedjtoldetermine their; consistency l                                    witMinputltoidesign calculationsgJtessured: core:and process [ parameters:are[ evaluated 2toidetermineithatithey are1within!theiboundslofithe:safetytanalysisEand safety analysi s1 cal cul ationalinodel s f areireviewed [to;verifyj that theydare adequateiforjrepresentation:ofithe; core conditionsMTheltequjredLCompletionjTime of;7;dayslis based;op;the;]ow; probability;of;a:DBA1 occurring lduring this! period aand; allows sufficient; time;;tolassessithe physical;conditjonl of;thelreactor;and;compl ete] any required; evaluations:ofuthe: core: design and: safety ana]ysesl CPSES Description of Changes - CTS 3M.]              6                                 7/29/98

IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 17 , 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS there are no required actions; thus, LC0 3.0.3 would be entered. LC0 3.0.3 requires that, within 1 hour, actions be initiated to place the plant in a condition in which the LC0 did not apply. Because this particular Surveillance R The proposed activity l would add required actions if the overall core reactivity balance was not within ilt ak/k of the predicted values. In the current Technical Specifications, equirement is only required in Modes 1 and 2. LC0 3.0.3 would further require that the plant be placed in HOT STANDBY (Mode 3) within the followino 6 hours. The proposed change, consistent with NUREG 1431, Revision mod.1fied;by. TSTF-1 g,wn f.daysJo evaluate the nSafety A[aiyw d octablich Fvpriate "gould allow M n -3.1 003 vpua W uw restrictions and/or sur"M' h g requirements. If these activities were not completed within t 72 hour 7' period, then the plant would be  : placed in Mode 3 within the fo hours. The requirement to periodically compare the measured and predicted overall core reactivity balances provides assurance that the analytical predictions upon which the  ; safety analyses are based accurate represent the actual core response. Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequata fne rancasantation of the core conditions. The required ._ Completion Time ofkEheues-7 days)s based on the low probability of a DBA TR 3.1 003 occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete any required evaluations of the core design and safety analyses. Following evaluations of the measurement, the core design and the safety analysis, the cause of the reactivity anomaly may be resolved. If it is concluded that the reactor core is acceptable for continued operation, then the predicted core reactivity balance may be renormalized and power operation may continue. If operational restriction or , additional SRs are necessary to ensure the reactor core is acceptable for continued ' operation, then they must be defined. The required Completion Time of 72 hours 7 days 's adequate for preparing TR.3'1 003 whatever operating restrictions - ail-_ that may be required to allow continued reactor operation. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below: 1 CPSES No SignifIcant Ha:ards Consideration - CIS 3N.] 48 7/19/98

l l IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS i NSHC LS 17 (continued)

                    "The Cwmission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility

licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the faci 1ity in accordance with the proposed amendment would not

1. Involve a significant increase in the probabi11ty or consequences of an l accident previously evaluated; or j 2. Create the possibility of a new or different kind of accident frorn any 1

accident previously evaluated; or

3. Involve a significant reduction in a margin of safety." ,

l I The following evaluation is provided for the three categories of the significant hazards consideration standards: i

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

l The proposed change does not involve any new operating activities or hardware changes: thus, the proposed changes has no effect on the probability of an accident. Although a small effect, the proposed change may slightly reduce the

probability of an accident by allowing additional time to resolve discrepancies, and thus avoid an unnecessary plant transient (shutdown).

l l Satisfaction of the Surveillance Requirement acceptance criterion provides assurance that the core related reactivity parameters used in the safety analyses adequately represent the actual core conditions. During the 72 hour l window following an initial failure of the Surveillance Requirement acceptance ! criterion, actions are established which would ensure continued agreement between the safety analyses and the actual core conditions, thereby maintaining the validity of the safety analyses. Therefore, there is no effect on the consequences of an accident previously evaluated. Because the available time is increased from 72 hours 7fdah j the probability of an accident occurring during tne 1.ime perivu wnen the - f plant condition is under review is slightly increased; however, the _ E 31003 l increase is small and has been previously found to be acceptable bv the NRC staff through the approval of NUREG 1431. Revision T

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

Operation for a period of time with a discrepancy between the measured and predicted core reactivity balances is allowed by the current Technical l Specifications: therefore, there is no possibility of a new or different kind i of accident. l CPSES No Significant Ha:ards Consideration - CTS 3N.1 19 7D9/98

Industry Travelers Applicable to Section 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF 9, Incorporated 3,1-1 NRC approved. Revision 1 TSTF 10 Not incorporated NA NRC rcjceted. ( P 3.1 002 Revisica 1 TSTF 11, Rcvision 1 Not incorporated --NA NRC rejccted. TP 3.1 002 TSTF-12, Incorporated 3.1 15 NRC approved. ITS Special Revision 1 Test Exception 3.1.10 is retained and re numbered as 3.1.8 consistent with this traveler and TSTF-136. TSTF 13, Incorporated 3.1 4 NRC approved. Revision 1 TSTF 14 Incorporated 3.1-13 NRCiapproyed. with ."00E 2 (P.3.1-002 Revision --34 retained in ITS 3 1.0 A ~1J..kJ1/6i, ove > - ' ' i V . 1R 3.1 005 TSTF-15. Incorporated NA NRClapproved,; Net W ep N [( P 3.1 002 Revision 1 gcncrically. TSTF-89 Incorporated 3.1-8 NRC approved. TSTF 107 Incorporated 3.1 6 NRC requested changes. c p.3.1 002 TSTF-108 Net Incorporated NA Net NRC approved. W ,g.3,1.coi RevisionL1 3.1-21 traveler cutoff date T5TF 110, Incorporated 3.1-10 NRC approved.- ,a.3,1.oo4 RevisionE23 MTT *c cnhance traveler

                                                                                 -justii, cation pcr NRC          ( p.3.1 002 ee m ente i

TSTF-136 Incorporated 3.1 9,"3 tlil5 73,1002 1 l l TSTF 141 Not incorporated NA Disagree with change; traveler e snart a % r cut off date. _ [ TSTF 142 Net-Incorporated 3.1'-22_NA NRC: approved.1Travelcr issuec ag ,1,003 I after cut-off date. Y ~

                                                                                                                              )

L---_------------_--------------------.-.--- --

1 1 Core Reactivity ~ 3.1-94 ) 3-1-33,1;2 3.1 REACTIVITY CONTROL SYSTEMS I l 3-1-33;1:2 Core Reactivity 1 i LC0 3-1-3 3i1.2 . The measured core reactivity shall be within i 1% ak/k of predicted values. 1 i APPLICABILITY: H0 DES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Measured core reactivity A.1 Re-evaluate core design and 72 hours , not within limit, safety analysis, and 7/ days 3.1-22 u 3.1 003 determine that the reactor core is acceptable for continued operation. AND A.2 Establish appropriate ating restrictions and I ([3ff003 B. Required Action and B.1 Be in H0DE 3. 6 hours associated Completion Time not met. l CPSESMark-sty ofNUREG-1431 - ITS3.1 ' 1-2

                                                                                                                    .                                     7/29/98

Core Re. activity SBN-(, s 200'F B 3.1.2 n.3,1 5 BASES ACTIONS A.1 and A.2 (continued) core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation a' th : r g ditions. The required Completion Time 7Edays 72 y 3 1s cased on the TR 3.1 003 low probability of a DBA ocedi. i% um ing this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anoc.aly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined. I TR 3.1 003 The required Completion Time okays 72 h' curs s adequate for preparing whatever operating rem u.uuns or Surveillance ) that may be required to allow continued reactor operation. fL1 If the core reactivity cannot be restored to within the it ok/k limit, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to , at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by 2 3.1.1.1 LC0 3.1;1lRequi_ red Action _A.1;would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without , challenging plant systems. I L CPSESMark-up ofNUREG-1431 Bases - ITS 3.1 B 3.1-12 S/1S/97

CHANGE NUMBER JUSTIFICATION 3.1 18 A MODE change restriction has been added to ITS 3.1.1. in the LC0 applicability, per the matrix discussed in CN 102 LS-1 of the 3.0 package. 3.1 19 Not Used. , 1 3.1 20 Consistent with current TS 3/4.10.3. " Physics Tests," ITS LC0 3.1.8 and its Condition C and SR 3.1.8.2 are modified to refer to " operating" RCS loops. Adopting the current TS is , acceptable since valid Tavg measurements are not obtainable for a L non operating loop. )

                                                                                                                                                                                         ~

311 21 ThelISTSTSR 3.1;8]llrequirementsto; perform'aLCHANNEL rR 3.1-001 OPERATIONAlgESTi(C0T)l.onltheiintermediatetand1powetrange  ! NIS channels:~ w ithinll12 hoursiprior to;initiatingzPtiVSICS TESTS;isfevised,tol delete thelphraseWithin 121 hours;" COT testing;is performed lonitheseichannelsl: prior to; reactor.;startup per LC0;3i3.1 G This change 21s consistent.1w1_th;TSTFt108;l 311i22 The; Completion'. Times.lfor.iITS13:1.2LRequired; Actions ~ A.1lland A;2;areiincreased;from;72ihourslto171daysLconsistent with W traveler,TSTF 142; w l CPSESDifferencesfrom NUREG-1431 -ITS3.1 4 7/29/98 e___-__-_____-_______

Y N Y Y Y K E E R C

                                                                          ~

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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR-3.1-004 thru TR-3.1-006 APPLICABILITY: CA, CP, DC, WC REQUEST: Revise Traveler Status Sheet to reflect NRC approval and latest revision number of travelers TSTF-14 Rev. 4, TSTF-110 Rev. 2, and TSTF-136. Change WOG-73 Rev. 1 to TSTF-234 (still under NRC review). Remove travele revision numbers everywhere except on the Traveler Status Sheet. There are no changes involved to any CTS mark-ups, ITS mark-ups, DOCS, or JFDs. ATTACHED PAGES: Enci 3A 3, 4,10 Encl 3B 1,2,7,8 Enci 4 25 EnctSA traveler pages (2) EncI 6A 2, 3 Enci 6B 1, 2 l l l [ l I l 1 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ _ _ _ . - -- )

1 CHANGE l NUMBER NSHG DESCRIPTION traveler TSTF 136. Traveler TSTF-9 relocated values for SDM to the COLR Rev. 'e 1. 'm_oved the only difference between ISTS LC0 3.1.1 and ISTS i LC0 3.1.2. Differences above and below 200*F will be addressed in the COLR. 03 01 A The footnote referring to Special Test Exceptions would be i deleted. This is acceptable because the requirements for Special Tert Exceptions are provided in separate LCOs.. ] 1 Therefore, a separate reference in the footnote is l redundant. 1 03 02 LS 3 Action Statement A.1 would be revised to require achieving j Mode 2 with k,,, < 1.0 instead of achieving HOT STANDBY if l the BOL HTC limit is exceeded and revised rod withdrawal ' limits cannot be established. This change corrects the discrepancy between the BOL Applicability and the ACTION, while ensuring that the plant is taken to a condition in which the LC0 is not applicable. Revising the current TS, albeit to correct an inconsistency, represents a relaxation in ACTION Statement a.1. 1 03 03 A The statement that administrative withdrawal limits required to meet Action Statement a.1 are in addition to insertion limits of another specification would be removed. This change is an administrative change because the statement is redundant to the requirements of Specification 3.1.3.6 and therefore can be deleted. 03 04 LG The requireemnt of current TS Action Statement a.2 which provides the detailes of how to verify theat HTC has been restored to within limits (i.e., calculation) for the all rods withdrawn condition prior to exiting Action Statement , a.1. is adressed in the ITS 3.1.3 Bases, j 03 05 TR 2 The requirement to submit a Special Report to the NRC would be deleted. This is in conformance with the ISTS. 03-06 LS 4 This change would incorporate a Note from ITS 3.1.3 allowing suspension of MTC testing near the end of the cycle when further significant changes to the HTC would not occur and result in exceeding the E0L limit. This , represents a relaxation in performing the surveillance l requirement. t 1 03 07 LG Not applicable to CPSES. See Conversion Comparison Table , (Enclosure 3B). l 1 i CPSES Description of Changes - CTS 3M.] 3 7D9/98 l J

CHANGE NUMBER HSliC DESCRIPTION 04 01 LS 5 This proposed change would make two changes to the Action Statement. First, it would alter the Action Statement shutdown requirement time limit from a combination of i 15 minutes to restore T.,, to within limits followed by 15 minutes to be in Mode 3 if T.,, could not be restored, to a single 30 minute limit to exit the Applicability if T.,, was not within its limit. Second, the Action Statement would be revised to require achieving Mode 2 with k,,, < 1.0 instead of achieving HOT STANDBY if the LC0 were not met (refer to TSTF 26). Regarding the first change, both the current requirement and the ITS requirement are essentially equivalent in that the plant is now required to shutdown and exit the applicability of the specification within 30 minutes after discovering that a parameter is not within its limits, if the parameter is not restored to within its limits in that 30-minute time period. If the LC0 can be satisfied at any time during the 30 minute time frame, the plant shutdown can be terminated. Regarding the second change, it represents a relaxation in current Action Statement requirements for plant shutdown, consistent with exiting the LC0's Applicability. 04 02 LS-6 The proposed change would revise the Surveillance Requirement for verifying that Reactor Coolant System (RCS) temperature (T,,,) is within limits by changing the Frequency to once per 12 in accordance with - industry traveler TSTF 2 Rev. 2. Ine current -- m 3.1 006 frequency requirements were nn 15 minutes prior to achieving reactor criticality which is redundant and unnecessary since Tavg must be within limits prior to entering the LC0 applicability, and at least once per 30 minutes when the reactor is critical and the (T,,, T,,,) l Deviation Alarm is not reset. The RCS temperature is maintained within limit: (1) to assure that the Moderator Temperature Coefficient is within the limits assumed in the accident analyses, (2) to assure that the neutron , detectors are not adversely affected by neutron I attenuation caused by low coolant temperature, (3) to j assure that the RCS and pressurizer response to thermal-hydraulic transients is as predicted, and (4) to assure that the reactor vessel temperature is above the nil-ductility transition reference temperature. The plant design incorporates monitoring of T,,, and j provides an alarm, the (T,,, T,,,) Deviation Alarm, as T,,, approaches its limit. This alarm condition requires a response by the operating staff. Therefore, at any time CPSES Description of Changes - CTS 3N.! 4 7/29/98 i

f CHANGE NUMBER HSt!C DESCRIPTION the public health and safety. Therefore, moving TR 3.1 004 this detail is ac. ' ab and is consistent with traveler TSTF-1 /. ncv. 1. 12 17 A Editorial changes made for clarity. Untrippable rods are i addressed through Action a: hence, there is not additional need to exclude those rods from these required actions. ' i 12 18 LG The technical contents of the Action Statement which allows continued power operation with a misa11gned rod are moved to the Bases for ITS LC0 3.1.4, Action B.1. 12 19 LS 18 Consistent with NUREG 1431, Rev. 1, the frequency at which the rod motion surveillance is performed is extended from 31 to 92 days. Thelproposedjchange;would 0-3 1 17 decrea.seltheitestingifrequency;at;Whichleachirod l cl ustericontro13 assembly;j s;exerci sod ito; demonstrate l theJab111ty;.otthe;; rods _,to;beltrippedj2 Current] Technical l Specifiestion) Surveillance 1 Requirement;[4;1;3 1;2] exerci ses;each ; rod;cl ustard controE assemblyiatij ea.stiten l stepsito; demonstrate 1thelability;ojltheirods3tolbe trfppedn$ This! surveillance ! requirementii sl performed iat leastlogceiper331; days Eger,1fying ;eackcontrolirod ii s i OPERABLE 1woulditequirelthatieach; rod beltripped;ZHowever? inMODES11EandL2;7 tripping;[eachIcontrollrod;wouldl result 4 l_ iniradiaMornaxial]poweritiltst.;orsoscillationsQ l l Exercising;eachlindividualicontrolirod;every 92 days Provides3jncreased; confidence 4that allfrodsicontinuelto;be OPERABLE 31thoutlexceedingithela]ignment:11mitteven;1f theylareinot 2egularlyitrj pped EMoving ieach ;controlirod byf101 steps;w1111not;Leause: radiaEorlaxial; power 2t11 tsGor oscill ations gtoloccurMThe; 92; day; Frequency 1 takes! 10to consjderatioM otherij gformationf avail ableitoltheloperatot inithe; controliroom;andf SR;311^(4;1';11whichj i s3 performed morelfrequentlyf andladdsito ;theldetermination;;of OPERABILITY [of 1 the: rod _sClBetween;ordduring fequired performances;oCSR;371;4 :2; (determinate onlofacontrol; rod OPERABILITY;by; movement)EjflaLeontrol2 rod (s)l1s

di scovered Lto;be,;junovable g butirenainsitrippableEthe l contro1Irod(s)11srcon_sidered;to;be10PERABLEiuntilithe surveill ancelintervalsexpj resMAtlany;timeZj fia icontrol tod(s)MsMemoyab.leEa1determinationTof;theLtrippability

                                                                      .( OPERABILITY)lofithelcontrollrod(s)1mustibe;madegand appropriate; action;takenMThrdecreaseMnfthe testing frequency;tojat;1eastionce;per;92;daysMs(recommended by Generic ;LetterJ93105 EItem ;4; 2 ;3 Control; Rod: Movement I                                                                      Testy i

CPSES Description of Changes - CTS 3M.1 10 7/29/98 I l i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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1 IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERAT10NS NSHC LS 6 I 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS m The proposed chan ould revise the Surveillance Requirement for verifying that Reactor Cool ystem (RCS) temperature (T.,,) is within limits by changing the Fr once per 12 hours in accordance with industry traveler TSTF 27.

           . 2. The current frequency requirements are within 15 minutes prior to                m 006 ng reactor criticality, which is redundant and unnecessary since T,,, must be within its limit prior to entering the LC0 Applicability, and at least once per 30 minutes when the reactor is critical and the (T,,, T,,,) Deviation Alarm is not              '

reset. The RCS temperature is maintained within limit: (1) to assure that the Moderator Temperature Coefficient is within the limits assumed in the acident analyses, (2) to assure that the neutron detectors are not adversely affected by neutron attenuation caused by low coolant temperature, (3) to assure that the reactor coolant system and pressurizer response to thermal-hydraulic transients is as predicted, and (4) to assure that the reactor vessel temperature is above the nil-ductility transition reference temperature. The plant design incorporates monitoring of T,,, and provides an alarm, the (T.,, T,,,) Deviation Alarm, as T ,, approaches its limit. This alarm condition requires a ' response by the operating staff. Therefore, at any time that T,,, is approaching its limiting value, the plant operators would receive an alarm and initiate corrective action. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:

                "The Corrinission maj make a final determination, pursuant to the procedures in 50.91 that a proposed amendment to an operatwg license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendrnent would not:

1. Involve a significant increase in the probability or consequences of an l accident previously evaluated; or 1

2. Create the possibility of a new or different kind of accident fran any accident previously evaluated: or

3. Involve a significant reduction in a margin of safety."

The following evaluation is provided for the three categories of the significant j hazards consideration standards: a i CPSES No Significant Hazards Consideration - CTS 3/4.1 25 7/29/98

Industry Travelers Applicable to Section 3.1 1 ( TRAVELER # STATUS DIFFERENCE _f COMMENTS TSTF 9, Incorporated 3.11 NRC approved. Revision 1 TSTI-10 Not incorporated NA NRC rcjccted. ( P-3.1 002 Reds 4en-1 TSTF 11, Not incorporated --NA NRC rcjected. (P.3.1 002 Revision i TSTF 12, Incorporated 3.1-15 NRC approved. ITS Special Revision 1 Test Exception 3.1.10 is retained and re-numbered as 3.1.8, consistent with this traveler and TSTF-136. TSTF 13 Incorporated 3.1 4 NRC approved. Revision 1 TSTF-14. Incorporated 3.1-13 pproved. with tiOOC 2 cp.3.1-002 l Revision h.sincd6..TS..0

                                                   ,__u-o,                      N
                                                  evii w ui' D                   1 R-3.1005 l TSTF 15      Incorporated            NA         NRC1 approved,(Not yct approve  c p.3 1 002 Revision 1                                       scncrically.

TSTF 89 Incorporated 3.1 8 NRC approved. TSTF-107' Incorporated 3.1-6 NRC rcquested changes. c 3.1-15 Revisionf1 TSTF 108 Net Incorporated NA Net NRC approved. es-ef 1R 3.1 001 Revision:1 3:1 21 traveler cutoff date. TSTF 110 Incorporated 3.1 10 IRC.approvedb 4R.3'1 004 Revisio 2:;. Yr tc' cn'ia(ce travelee justificat ca per iPS cp.3.1 002 een.sents-TSTF 136 Incorporated 3.1 9 J3.'l 15 c p.3,1 002 TSTF 141 Not incorporated NA Disagree with change: traveler l issued after cut-off date. I

TSTF 142 Het-Incorporated 3;1!.22..NA NRC-':approvedrl;Travcler i;sucm ,g.3,1.oo3 after cut-off date mei E 73 ~ ^^" ' 1 Incorporated 3.1 7 ( P 3.1 002 F-234

                                                                                                                              .__. lR 3.1 006 l

WOG 105 Incorporated 3.1-16 I i l I l l l l i l l

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I j CHANGE NUMBER JUSTIFICATION 3.1 6 ITS LC0 3.1.4 would be split into two separate statements to clarify that the alignment limit is separate from OPERABILITY of the control rod. The CONDITION A wording is broadened from "untrippable" to " inoperable" to ensure the CONDITION encompasses all causes of inoperability. Previous wording was ambiguous for rods that, for instance, had slow drop times but were still trippable. These slow rods are inoperable rods, and the change clarifies the appropriate ACTIONS. The Bases are changed to reflect the changes to the LC0 and CONDITION A. These changes are based on traveler TSTF 107. 3.1 7 This change to the ISTS would incorporate into ITS LC0 3.1.7. an Action Statement that was previously approved as part of the Callaway and Wolf Creek licensing basis as revised in Enclosure 2. The Action Statement would permit continued POWER OPERATION for up to 24 hours with more than one Digital Rod Position Indicator per rod group inoperable. The Action Statement specifies additional required actions beyond those applicable to the condition of one DRPI per group inoperable. The Bases for this change also would be incorporated into the Bases for the niant ITS. These ~ changes are consistent with traveler (STF 234' '400 73. Ph

                          -1. The Note under the ACTIONS is changed to be consistent         TR-3.1 006 with the new Required Actions.

3.1 8 The Frequency for ITS SR 3.1.7.1 for comparing DRPI and group demand position would be changed from 18 Months to "Once prior to criticality after each removal of the reactor vessel head." This change makes it clear that the surveillance must be performed each time the head is removed and that it is not tied to an absolute time interval. This change is based on traveler TSTF 89. 3.1-9 This change would eliminate ISTS 3.1.2 because the SDM requirements for MODE S have been incorporated into Specificat .1.1 in accordance with TSTF 136. Traveler i TSTF 9. Pa.q' relocated values for SDM to the COLR which TR 3.1 006 l removed t e only difference between 151b LCO 3.1.1 and1TSI j LC0 3.1.2. Differences above and below 200*F will be addressed in the COLR. Subsequent sections have been re-numbered. 3.1 10 Several surveillance (e.g., rod position deviation monitor 7g.3,1.go4 and rod insertion limit monitor in this section) contain actions in the form of increased surveillance frequency to be performed in the event of inoperable alarms. These actions are moved from the TS to licensee controlled documents since the alarms do not themselves directly relate to the limits. This detail is not required to be in the TS to provide adequate l protection of the public health and safety. Therefore, moving this detail s acceptable and is consistent with traveler TSTF-110 ^ -" ' CPSES Differencesfrom NUREG-1431 -ITS 3.1 2 7/29/98

1 i CHANGE NUMBER JUSTIFICATION 3.1-11 Not Used. 3.1 12 The Required Actions for inoperable DRPI in ITS 3.1.7 are revised per the current licensing basis to note that the use of movable incore detectors for rod position verification is an indirect assessment at best. The position of some tods can not be ascertained by this method.  ! 3.1 13 This change adds an LC0 requirement and SR to MODE 2 Physics Tests Exceptions 3.1.8 to verify that thermal power is less than or equal to 5 percent RTP. The LC0 requirement and SR were added to verify that THERMAL POWER is within the defined power level for MODE 2 during the performance of Physics Tests. since there is an Action that addresses THERMAL POWER not within limit yet there was no corresponding LC0 or surveillance requirement. The Surveillance Frequency of 1 hour is retained from the current TS. This change is based on TSTF-1 visicr, TR 3.1 005 3.1 14 Not used. 3.1 15 Consistent with TSTF 12 Revision 1, ISTS LCOs 3.1.9 and 3.1.11 are deleted. The physics tests contained in LC0 3.1.9 were only contained in some initial plant startup testing programs. The physic test exception can be deleted since these physics tests are never performed during post refueling outages. The physics test that LCO 3.1.11 required was the Rod Worth Measurement in the N-1 condition. The use of other rod worth measurement techniques will maintain the shutdown margin during the entire measurement process and still provide the necessary physics data verification. Since the N 1 measurement technique is no longer used, the SDM test exception can be deleted. This change and traveler TSTF-136 renumbers ISTS 3.1.10 to ITS 3.1.8. 3.1-16 This change adds the requirement to perform SR 3.2.1.2 in addition to SR 3.2.1.1 during performance of ITS 3.1.4 Required Action B.2.4. The intent of Required Action B.2.4 is to verify that Fo(Z) is within its limit. Fa(Z) is c?proximated by F8(Z) (which is obtained via SR 3.2.1.1) and F#(Z) (which is obtained via SR 3.2.1.2). Thus, both Fj(Z) snd F#(Z) must be established to verify Fa(Z) . This change is consistent with traveler WOG 105. 3.1 l'/ Consistent with current TS LC0 3.1.3.2, ITS 3.1.7 Condition C is j , clarified to state that the inoperable position indicators are i inoperable DRPIs. 3.1-18 A MODE change restriction has been added to ITS 3.1.1. in the LC0 applicability, per the matrix discussed in CN 1-02 LS-1 of the 3.0 package. LC0f3.0.4~has.been1 revised so that changes 3.1 27 in _ MODES Lotother specifled ; conditions 11n1the_ Applicability l that are:partfofjafshutdown of;the~ unitLshallinot be CPSESDifferencesfrom NUREG-14.11 -ITS3.1 3 7/29/98

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Attachment 2 to TXX-98182 Page 1 of 4

                                                                                                                                                          )

i 1 JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS CTS 3/4.2 - POWER DISTRIBUTION SYSTEMS ITS 3.2 - POWER DISTRIBUTION SYSTEMS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES i l l 1 l

Attachment 2 to TXX-98182 Page 2 of 4 f f I INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED l NUMBER I I 3.2.G-1 DC, CP, WC, CA YES ! 3.2-1 CP YES i 3.2-2 DC NA 3.2-3 DC, CP, WC, CA YES 3.2-4 DC, CP, WC, CA YES 3.2-5 CA NA 3.2-6 DC, CP, WC, CA YES 3.2-7 WC, CA NA j 3.2-8 WC NA J 3.2-9 WC NA { 3.2-10 DC, CP, WC, CA YES j CA 3.2-001 DC, WC, CA NA j CA 3.2-002 DC, WC, CA NA ' CP 3.2-ED CP YES CP 3.2-001 DC, CP, WC, CA YES DC ALL-005 (3.2 changes only) DC NA DC 3.2-ED DC NA DC 3.2-001 DC NA

                                                                         . TR 3.2-004                                         DC, CP, WC, CA  YES WC 3.2-001                                    WC, CP          YES

Attachment 2 to TXX-98182 Page 3 of 4 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:

1. Each licensee is submitting a separate response for each section.

                ?. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

3. If a licensee initiated change does not apply, "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

4 The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A, 3B, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses tojustify thair inclusion in each submittal. In those cases, the response will be prefaced with a heading such as " PLANT SPECIFIC RESPONSE . ....."

5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.

6. A marginal note (the Additional Information Number from the index)is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.

7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one or more of the other licensees. These l changes are not included in the additional information for the licensee to which the change does not apply, as the changes are only for consistency, do not technically affect the request for that licensee, and are being provided in the additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment request supplement to be provided later.

i l L______----___________.__ - I

{ l f Attachment 2 to TXX-98182 Page 4 of 4 , i JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR l PROVIDING ADDITIONALINFORMATION ) l (continued)  ! l 8. The item numbers are formatted as follows: [ Source][lTS Section]-[nnn) l Source = Q - NRC Question CA- AmerenUE DC-PG&E WC - WCNOC CP - TU Electric TR - Traveler ] ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL"is used for the section number. j nnn = a three digit sequential number or ED (ED indicates editorial correction with no impact on meaning) 1 l

l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q3.2.G-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.2.x Bases General l I There have been a number of instances that the specific changes to the STS Bases are not- 1 properly identified with redline or strikeout marks. Comment: Perform an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal. 1 FLOG response: The submitted ITS Bases markups for Section 3.2 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorial in nature and would not have affected the review. Examples of editorial changes are:

1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced.

2) Changing a verb from singular to plural by adding an "s" without " redlining" the "s".

3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g.,

SR3.6.6A.7).

4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference.

5) in some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.

6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.

7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was determined to not be applicable, the text was then struck-out and remains in the ITS Bases mark-up.

Differences of the above editorial nature will not be Povided as attachments to this response. The pages requiring changes that are more than editorial and are not consistent with the markup methodology are attached.

                                                       ' ATTACHED PAGES:

Enci5B B 3.2-4 A sentence was inadvertently deleted and is added back B 3.2-7 Revised sentence was not accurately relined / struck-out Bracketed factors were not redlined B 3.2-8 The word " ensures" was deleted instead of being struck-out B 3.2-13 Bracketed "(Ref.1)" was not redlined B 3.2-14 The word " Limits" was incorrectly redlined l Bracketed "(Ref. 3)" was not redlined l B 3.2-16 The word " Required" was inadvertently deleted and is added back I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _J

Fa(Z) (Fa Methodology) B 3.2.1 BASES (continued) LC0 (continued) FaiZ) is increased by 3t to account for manufacturing tolerances and:further;1ncreasedjby 5*1to account for measurement a uncertainties. Fj(Z)11s,anl excellent . approximation ;for; Fo(Z) whenhfeacto is (at thetsteady state: power;at which:thejincore flux map 0 3.2m gasitaken! I The expression for F"a(Z) is: F"a(Z) = Fj(Z) W(Z) where W(Z) is a cycle dependent function that accounts for power distribution transients during normal operations. W(Z) is included in the COLR. The Fa(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200*F during either a large or small break LOCA. This LC0 requires operation within the bounds assumed in the i safety analyses. Calculations are perforr,cd in the core design l process to confirm that the core can bc controlicd in such c  ! anner during cperatien that it can stay within the LOC ^ F,(Z)

                                                                            +imitt.- If Fa(Z) cannot be maintained within the LCO limits,(a                 ,

reduction of the core power is required. l l Violating the LC0 limits for Fo(Z) may produces unacceptable l consequences if a design basis event occurs while Fa(Z) is i outside its specified limits. j If the power; distribution _measurementsfare: performed at a; power level 11ess;thanL100%_RTPHthan the E'a(Z) land F"a(Z)svaluesLthat woul_d: result; from: measurement _siif;the core;was,at1100* RTP should be inferred:from the_ available"information; lAicomparison'.of these inferred values with F/'P;assur,estcompliance with the LC0 at all power._ levels; APPLICABILITY The Fo(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses. Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power. (continued) CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-4 7/29/98

Fa(Z) (Fa Methodology) B 3.2.1 BASES (continued) modified, however, by one of the Frequency conditions that requires verification that Fj(Z) and Fl(Z) are within their specified limits after a power rise of more than -1M20% RTP over the THERMAL POWER at which they were last verified to be within specified limits. Because Fj(Z) and F#(Z) could not have previously been measured in this foria reload core, there is a l second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of Fj(Z) and F#(Z) are made at a lower power level at which adequate margin is j available before going to 100% RTP. Also, this Frequency i condition, together with the Frequency condition requiring verification of Fj(Z) and F#(Z) following a power increase of more than -1M 20%, ensures that they are verified within'24 h rs f rom when equilibrium, condition _s are(aEhieved a - 0-3.2.G 1 s^^n QTP (or any other level for extended operation)  ; achicved 1 Equilibrium; conditions are achieved;when the l core,is,sufficiently; stable;such thatJthe uncertainty 1 allowances associated.with the;measur_ement are,vali_dj In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of Fj(Z) and F#(Z) . The Frequency condition is not intended to require verification of these parameters after every -1M 20% increase in i power level above the last verification. It only requires verification after a power level is achieved for extended operation that is -1M 20% higher than that power at which Fo was last measured. SR 3.2.1.1 Verification that Fj(Z) is within its specified limits involves j increasing F5(Z) to allow for manufacturing tolerance and i measurement uncertainties in order to obtain Fj(Z). ) Specifically, Ff(Z) is the measured value of Fo(Z) obtained from incore flux map results and Fj(Z) = F"c(Z) [1.0815]

                            ) 1.                                       .             Fj(Z) is then compa                                                                         0 3.2.G 1 The limit with which Fj(Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) j                        provided in the COLR.

l (continued) CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2 7 7/29/98 i

Fa(Z) (Fa Methodology) B 3.2.1

     . BASES (continued) meeting;the7100%;RTP'F                       a (Z) 11miti provides assuranc that the Fj(Z) limit is met when RTP is achieved, because                                       0 3.2.G 1 peaking factors generally decrease as power level is increased.

If THERMAL POWER has been increased by 2 iM 20% RTP since the last determination of Fj(Z), another evaluation of this factor 1<, required E12-} 24 hours after achieving equilibrium conditions at this higher power level (to ensure that Fj(Z) values are being reduced sufficiently with power increase to stay within the LC0 limits). The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS). SR 3.2.1.2 The nuclear design process includes calculations perfor;;;cd to deter =ir.c that the cere car. 50 cperated "ithir. the F,(Z) liritt. Because flux maps are taken in steady ; tate equilibrium conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, Fj(Z), by W(Z) gives the maximum Fa(Z) calculated to occur in normal operation, F#(Z). The limit with which Fl(Z) is compared varies inversely with power and directly with the function K(Z) provided in the COLR. The W(Z) curve is provided in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. Fl(Z) evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive: and l

b. Upper core region, from 85 to 100% inclusive.

(continued) l CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-8 7/29/98 < I

Flu B 3.2.2 BASES correlation. All DNB limited transient events are assumed to beginwithanFlu value that satisfies the LC0 requirements. Operation outside the LC0 limits may produce unacceptable consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant. APPLICABLE Limits on Flu preclude core power distributions that exceed SAFETY ANALYSES the following fuel design limits:

a. For ANS Condition II events, there must be at least 95%

probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition:

b. During a large break loss of coolant accident (LOCA), peak claddin, temperature (PCT) must not exceed 2200*F:

c. During an ej: ced rod accident, the energy deposition to the fuc12av.er@elfuelspe11.etlenthalov at1the; hot spot must not exceed 280 cal /gmQand - Q 3.2.G 1

d. Fuel design limits required by GDC 26 (Ref. 2) for the i condition when control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn.

For transients that acy bc ONS limited, the Rcactor Occiant Syste #1cw =d P, are the corc par;;;ters of aest import nce. ThelimitsonFlu ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the rinimum DNBR to the 95/95 DNB criterion of [1.3] using the ["3] C!!F applicable.to;.a ;specificiDNBR correlation. This value provides a high degree of assurance that j the hottest fuel rod in the core does not experience a DNB condition. The allowable Flu limit increases with decreasing power level. i This functionality in Flu is included in the analyses that (continued) l CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-13 7/29/98

Flu j B 3.2.2 ' BASES provide the Reactor Core Safet (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the 0-3.2.G-1 core limits is modeled implicitly use this variable value of Flu in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial Flu as a function of power level defined by the COLR limit equation. 1 The LOCA safety analysis indirectly models also;uses Flu as an input parameter. The Nuclear Heat Flux Hot Channel Factor (Fa(Z)) and the axial peaking factors are inserted directly into 1 the LOCA safety analyses that verify the a - ility of the resulting peak cladding temperature Ref.T3) . 0 3.2.G 1 The fuel is protected in part by compliance withjTechnical Specifications which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The l following LCOs ensure this: LC0 3.2.3, " AXIAL FLUX DIFFERENCE (AFD)," LC0 3.2.4, "OUADRANT POWER TILT RATIO (0PTR)," LC0 3.1.7.

                                                            " Control Bank Insertion Limits," LC0 3.2.2 " Nuclear Enthalpy Rise Hot Channel Factor (Flu)." and LC0 3.2.1. " Heat Flux Hot Channel Factor (Fa(Z))."

Flu and Fa(Z) are measured periodically using the movable incore detector system. Measurements are generally taken with the core at, or near, steady state equilibrium CP 3.2.ED conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD. OPTR. and Bank Insertion Limits. Flu satisfies Criterion 2 of the NP,0 Policy Statement 10CFR50;36(c)(2)(11). LC0 Flu shall be maintained within the limits of the rcisi.ionship provided in the COLR. The Flu limit is _ representative of idcatifice the coolant flow channel with the maximum enthalpy rise. 'his channel has the l least heat removal capability and thus the highest probability for a DNB condition.

(continued)

L CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.214 7/29/98

Fin B 3.2.2 BASES Fin to within its. limits without allowing the plant to remain in an unacceptable condition for an extended period of time.EThe restoration of;the; peaking; factor;to;within;1ts111mitsiby1 power reduction or;contro11 rod; movement [doesinotgestore[ compliance with;the:LCOEHhus;1thisfconditionican;not;be; exited;unt1Ra yal id 'surve111 ance2 demonstrates 1 comp 11anceiwithithe;LC0J Condition A is modified by a Kote that requires that Required Actions A.2 and A.3 must be completed whenever Condition A is entered. Thus, if p;nr i; not reduced b;;;;;; even]jfjthis Required Action is completed within the 4 hour time period, Required Action A.2 ncverthcic;; requires another measurement and calculation of Fin within 24 hours in accordance with SR 3.2.2.1. llcavcr. if pc.:;r i; reduced bals'.; 50 RTPhequir A.3 requires that another determination of Pau must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours after reaching or exceeding 95% RTPE however) THERMAL POWER;does;not;have;to be;reduceditofcomply;withithese requirements. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours. A.1.2.1 and A.1.2.2 If the value of Flu is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to

                        < 50% RTP in accordance with Required Action A.1.2.1 and reduce the Power Range Neutron Flux-High to s 55% RTP in accordance with Required Action A.1.2.2. Reducing RTP powerfto < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours for Required Action A.1.2.1 is

( consistent with those allowed for in Required Action A.1.1 and provides an acceptable time to reach the required power level

                       - from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours for Required Actions A.1.1 and A.1.2.1 are not additive.

The allowed Completion Time of 8 72 hours to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints:t1however! CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-16 7/29/98

j l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.2-1 APPLICABILITY: CP REQUEST: ITS 3.2.3 Axial Flux Difference CTS 3/4.2.1 Axial Flux Difference (Comanche Peak) DOC 01-05-M ITS Required Action D.1 > Comment: If the required power reduction, resulting from accumulated AFD penalty minutes, is not accomplished within the required completion time, the ITS had a note requiring a reduction in power to less than 15% RTP regardless whether the AFD was within limits. TSTF-112, Rev.1, deleted the note requiring the automatic reduction in power to < 15% RTP. Review and evaluate TSTF 112, Rev.1, and determine if this TSTF change can be adopted by Comanche Peak. FLOG Response: TSTF 112, Rev.1, is being adopted by Comanche Peak. ATTACHED PAGES: Enc! 2 3/4 2-2 Enci 3A 2 EnctSA Traveler page,3.2-11 Enci 5B B 3.2-26 Encl 6A 4 Encl 68 3 l

POWER DISTRIBUTION LIMIIS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued)

b. With the indicated AFD outside of the required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours or  ;

outside the Acceptable Operation Limits specified in the COLR and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER:

1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.JThisiaction;sust;belcompleted f E 05-M d whenever! action bEisienteredi end;0R;

2. Reduce thc Powcr Rangc Mcutron flux "igh Trip Satpcints 4'. g g" to isss than or aquel to 55 cf RATED TilEPJiAL P6WER within the acxt 4 hours.

[C(NEW)1 Reduce;TH[N NR to'less than 1Rr RTP within the(next 0105 H 9 houre d This cctic, ;ust.b; ;;,,,,,1cted.wh;r,;ycr c D 0 3.2 1 Gavc i: not3apictedMithiA00 :iouteb

                                                             -.                                                 /      -

c. With the indicated AFD outside of the required target band for more than I hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the required target band.

I SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AfD "onitor Alor; is OPEPELE, and

2) At least cace per hour for th first 24 hours after restoring the AfD "cnitor Alar; to OPEPELE status. 101 077LG i

b. "snitoring and icgging the indicated AFD for cach OPEPE LE cxccre chanaci at least cacc per hour for the first 24 hours and t01 07 LGL.

l st least once per 30 minutcs, when the AfD "cr.itor Alar; is inspciabic. The leggcd valucs of the indicated AfD shall be assu;cd to cxist during the ir.tarval preceding cach logging. I CPSESMark-up ofCTS 3N.2 3/42-2 July 29,1998

CHANGE NUMBER R$liC DESCRIPTION 01 05 H Consistent with NUREG 1431, additional requirements are imposed in the event reactor power is required to be reduced to less than or equal to 50% RTP due to accumulated AFD penalty minutes. In the current TS, if the required power reduction is not completed in the required time, LC0 3.0.3 would be entered. However, once the power was reduced to less than 50% RTP, the LC0 3.0.3 could be exited. Although not the intent, the required power reduction to 50% RTP c'+ N be extended over an additional hour. In the ISTS, if the required power reduction is not completed in the required time, the thermal power must be reduced to less than or ego. to 15% RTP over the next 9 hours, howeyer.lthisiaction requirement 1may_beiexitedyhen power 11s;reducedibelow 50%; Thi.s;is; appropriate 1because; operation;below 50% RTP;is notJconstrained bylAFDi Thcrc is ac opportunity to halt the powr decreasc prior to reaching 15 % itT". Thi; casures the applicability of th LC0 is exited in a timely snncr if thc LC0 cr the required actions cr; act ;;;;t. EIn addition, the requirement is added that action to reduce power below 50% RTP must be completed whenever the action is entered.-} 01 06 LS 2 With the Axial Flux Difference (AFD) outside of the required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours, or outside the Acceptable Operations Limits with THERMAL POWER between 90% RTP and 50% RTP. the current Technical Specifications required that the thermal power be reduced to less than 50% > i RTP and that the Power Range Neutron Flux High reactor trip setpoints be reduced to less than or equal to 55% RTP within , the next 4 hours. The requirement to reduce the high neutron { flux reactor trip setpoints is proposed to be deleted. The proposed change is considered to be acceptable because reducing the power level to < 50% RTP maintains the plant in a relatively benign condition where the axial flux distribution is not a significant accident analysis input. This change is consistent with NUREG 1431 and WOG letter 0G- l 90 54, 9/5/90. l I 01 07 LG As described in industry traveler TSTF 110 which modifies NUREG 1431 requirements, several surveillance contain actions (in the form of increased surveillance frequency) to l be performed in the event of inoperable alarms. These l l actions are moved from the TS to licensee controlled j documents. The alarms themselves do not directly relate to I the LC0 limits. This detail is not required to be in the TS { to provide adequate protection of the public health and j safety. Therefore, moving this detail is acceptable. ' l 01 08 Not used. I CPSES Description of Changes to CTS 3N.2 2 July 29,1998

I INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.2 TRAVELER # STATUS DIFFERENCE e COMMENTS l TSTF-24 Not Incorporated NA Not NRC approved as of traveler cut-off date. TSTF 95 Incorporated 3.2-06 Approved by NRC TSTF 97 Incorporated 3.2-07 Approved by NRC TSTF 98, Rev. I Incorporated 3.2-03 TSTF 99 Incorporated 3.2-08 Approved by NRC TSTF-109 Incorporated 3.2 15 Approved by NRC TSTF 110 Rev. 2 3 Incorporated 3.2 10 Approved by-NRC n' .2 004 TSTF 112. Rev. 1 ( Net Incorporated NA 3.2 20 Not NRC appr^ved 03 21 D

                   \                                      as of traveler ettt off date.

ll o Approved'by,NRC;f Applicable;to.,CAOC p1 ants. (CPSES"only)/ TSTF 136 Incorporated NA Approved by.NRC u. .2 004 TSTF 164 Incorporated 3.2 11 Applicable to CAOC only. (CPSES) L^C ^5. propc;cd Incorporated 3.2 05 a.3.26 Rev. 2 TSTF 241 3.2-109 Reval WOG 105 Incorporated 3.2 16 l l l I

AFD (CAOC METHODOLOGY) 3.2.3 CONDITION REQUIRED ACTION COMPLETION TIME

                 -         ~

NOTE D.1 Reduce THERMAL POWER to 9 hours D.y"a' quired Acticr. 0.1 must< 15% RTP. 3.2 20 be ce 9 cted wher.cVer 1 0-3.2-1 Cor.diticr. D is cr.tcred y w_ .J Kequired Action and associated Completion Time for Condition C not met. l 1 l l CPSESMark-up of NUREG-1431-ITS3.2 3.2-11 July 29,1998 _ _ _ _ . _- -___ ___ ____ - _-___ -__- - ~

AFD (CAOC Methodology) B 3.2.3 BASES (continued) operation limits.) The Completion lime of 30 minutes allows for a prompt, yet orderly, reduction i.i power. Condition C is modified by a Nate that requires that Required Action C.1 must be completed whenever this Condition is entered. D2.1 If Required Action C.1 is not completed within its required Completion Time of 30 manutes, the axial xenon distribution starts to become significantly skewed with the THERMAL POWER 2 50% RTP. In this situation, the assumption that a cumulative penalty deviation time of 1 hour or less during the previous 24 hours while the AFD is outside its target band is acceptable at < 50t RTP, is no longer valid. Reducing the power level to < 15% RTP within the Completion Time of 9 hours and complying with LC0 penalty deviation time requirements for subsequent increases in THERMAL POWER ensure that acceptable xenon conditions are restored. This "equired Actica must also bc impicm ated cither if the cumulative penalty deviation tiac is > I hour during the previous E4 hours, or the AfD is not within the. target band and not within the acceptable operatica limits. Condition 0 is modified by a Note that requires Action 0.1 0-3.2 1 bc completed when ver this Condition is entered. SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The AfD is monitor ^d on an automatic basis using the unit process eemputcr that has an AfD monitor alarm. The computer determies the 1 minute avcrage of cach of thc 0"E"""LE cxecrc detector outputs and provides an alarm message i- cdictcly if the ATOs for two or more 0"E",'s"LE cxecrc chenricis are cutside the target band

                                                   ..J     &L. Tlif*nLA A t nnt Pf"M       J.                         M.J-            ..&J       .6 Tt      L ussu bisw e s s6u u srw i VFULis         sa ' An&

m s u 's MTM s\ss uus isty upw s u w s ui s uk 11 C nnt.nr n 11. nn& nTn L..& m . se& nTn &L. ..__..&-- .._J,. . i VFvust swwwsa sum stia uub - c. A s ie iria , bssw vvissyu bu s awa sua use j alarm message whcn the cumulative pencity deviation time is

I hour in the previous 24 hours.

i (continued) I l CPSESMark-up of NUREG-1431 Bas ITS 3.2 B 3.2-26 July 29,1998

l CHANGE NUMBER JUSTIFICATION 3.2 15 This change incorporates industry traveler TSTF 109. Action A.2 would require the OPTR be determined rather than 0 3.2 10 performing a specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1. The;.no.tefandlFrequency;for;SRL3;2.4'.2"are reyised;toibeTconsistentiwith;typicalipresentation formats;that proyide:foCa'petiod;of;timelafter; establishing; conditions; The note for SP, 3.2.4.0 is changed to require perfor;;nce if sna "cr

                          ..~,m  w . .s . . .r ~ . . m ..~rs . ,s. These changes are acceptable because they clarify the ISTS regarding frequency and use of incore flux monitoring for OPTR measurement. The changes reflect that incore detectors provide an acceptable QPTR determination during all plant conditions.

3.2 16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that Fa(Z) is within its limit. Fa(Z) is approximated by Fj(Z) (which is obtained via SR 3.2.1.1) and F#lZ) (which is obtained via SR 3.2.1.2). Thus, both Fj(Z) and F#(Z) must be established to verify Fa(Z). This change is consistent with traveler W0G-105. 3.2-17 Not used. applicabic to CPSCS. Sec conycrsion compariscr. 0 3.2-3 tabic (cnclosurc SB). 3.2-18 Not applicable to CPSES. See conversion comparison table (enclosure 6B). 3.2 19 Not applicable to CPSES. See conversion comparison table (enclosure 6B). - = 3;2f20 This'changelincorporatesiTSTE112;(i;egidel.eteslthe; Notec in 0-3.2 1 l Condition D which;requiredithat required? action:D;1;be completediwhenever: Condition D'i.sjentered); IDeleting t.he

Note; permits;eziting;the;requirediaction when power 11sireduced l bel oW1500 RTM Exitingithelrequi red iaction j i s1appropri ate ..because operation ibel owl 500 R,TP11 sj noticonstra1ned ;with1 respect L to ' AFD; 1

CPSES Differencesfrom NUREG-1431 - ITS 3.2 4 7/29/98 L

a a P C m - s s s e e e o o o Y Y y N N N k e e r 2 C f 3 l o W m N s s s s s O Y e Y e y e e e o I Y Y N T C k 2 E a 1 S e 2 P 3

   -  e                                    ;

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• N N N N C

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                                                                                              ;o I      ts tf in2      .

m m n e D us cl 3 . a . e r uA 3 t t' i er m e0 3 r

                                                                                         ) ei ytt d eel s      F     ,oM t

sec0 0 R n l oc: nN ) R nr piR . a 4 t st c' S r o a. d ii us eS vv e  ;- ra c. n en r r scgeo

                                                                                          ;ed  e O       eqarr                      m2             o;      miec                en sh er tt re F      h e t rgsf .      uo ~         F o3 Wf      nT i   ea pi rt r c i

f o c % an ;s it u n n v s co i eoc i + l eqe E N s wiec eomnt . c n i o f c n r o ur cn re q d e ouy ; e r.s pt e; L O t nr oo . t i eoi eo' t msf Cl ii I a o n a t pt t rt si O et B T ro 2.f n m t c c o y eijAd a D'. pA ehh ra c ~ A P sA rdi R ,tb6 g ,C ( Ct 4, h  : I o c p Tk 1 ne n T R r on pt T . - d d n e f s = i c P edd1 a c O en c h n

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                                                                                       ;,eedi o

r c ; J c a g; et C cone . a r t vC N S ni a r; . t u i n

                                             - r ;

e eC co i n1 ri E i th o . . h spN cl ; a( id 0 ct mR .. n q - t f t ; i p 1un D S lyA r eP T ri c e e i R

                                             ^ a y r cf          soe" l n eW m:

nio(h t' f m c2 qo l e C, en em cc I l e sQ h s ci cl r R a .h i 9t a er i t0 a cnic u c c n r a o rf iwt e;p c c i f r r od pl i oiin s ai iT ca e mfEhi j r c A r1 reiu p a - b mqs t d r c ; h df ut e o qcmk eo eTh r cSi c r. P pFd r c n h e y t m 1. eAA et h p ; w o, _ M T e eeri eS nct t o r m c ; o n t. n e r d n rl eit ss

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_ O gTi n eoc n madt r b o f c . g nvT r

                                                                .S     ci C          f St a o_ E and C                                           ; ..d  ~ i

_ arrl c s r ae id f S roe h eel od e r p e e .d hRt u el eeP oit cl t it c ; r . c nqv ov s pt N eee g.t i c uh ec owl u C,r i le - O s vd vd r ec p u u n q e s c s1. rR r peseicnm o sod p i a eai i3 r I h reush n eh h c t.; h4 uoph eh h noo T tb s uci R w T a. T1 ct aT rt tic c S R E V - N O R _ C E 5 6 7 8 9 0 _ B 1 1 1 1 1 2 M - - . - _ 2 2 2 2 2

                                                                                       ,23 U

N 3 3 3 3 3

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.2-3 APPLICABILITY: CA, CP, DC, WC REQUEST: 11S 3.2.1 Heat Flux Hot Channel Factor CTS 3/4.2.2 Heat Flux Hot Channel Factor (All FLOG Plants) DOC 02-06-A l JFD 3.2-12 l ITS SR 3.2.1.1 & 3.2.1.2 Frequency l Comment: The ITS SR freqten v has been changed from the STS frequency of 12 hours to 24 hours. This is based upon the incorrect justification that the CTS would allow 24 hours based upon ITS SR 3.0.3, since the CTS does not specify a frequency. Adopt the STS SR frequency of 12 l hours. 1 FLOG Response: The change descriptions (DOC 2-06-A & JFD 3.2-12) will be revised to provide a basis for the 24 hours that is predicated on the time required to perform the surveillance. DOC 2-06-A is also been I revised to be DOC 2-06-M because this change is more restrictive than the CTS. I Callaway and Wolf Creek are incorporating this change (DOC 02-06-A, JFD 3.2-12) in lieu of ) maintaining CTS which did not specify any completion time. DOC 02-13-LG (applicable to Callaway only) and JFD 3.2-17 are no longer used. ATTACHED PAGES: Enct 2 3/4 2-6 Enci 3A 4, 5 Enct 3B 3, 4 Enci 6A 3, 4 EncI 6B 2, 3 l l l-

POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS (Continued)

e. Measuring eF 8 [/.(Z);and F/(Z)"according to the following schedule: 02-01'-LG 'l I NEW Oncela fteteach er.efuel ing; prior; toiexceedi ng f 75%; RTP;

                                                                                        < 02 05 H -

1. OpenWithinf24 hours _after achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL m

02 06-HA h POWER, the THERMAL POWER at which Fe(Z) was last o.3.2-3 i determined *, eefand

2. At least once per 31 Effective Full Power Days thereafter.

whichricr occurs first,

f. With measurements indicating mdximum F/(Z) mrr K(Z) has increased since the previous determination of F/(Z) either of the following actions shall be taken:

1) Increase F/(Z) by an allowance t-24-as specified in the COLR and verify that this value satisfies the :02-01 LG; relationship in Specification 4.2.2.2d, or

2) Fa"(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum F/(Z) ,

is not increasing.

                           ""?

K(Z) 9 Mth the relationships specificd in Specification 4.2.2.2d abcic not being satisfied: I B Calculate the percer,t that Fe m If F/(Z);. exceeds its limit by the fclicwing cxpression; 102 01 LG:

                        =ar,ime: F/(Z)x"(Z)
                                                  -1   x 100 for P r 0.5 0icr Z     F,"                                                           .

x K(Z) i ?

Y,imum F/(Z) x '.!(Z) 1 x 100 for P 2 0.5, x KCZ)

• Power level may be increased until the THERMAL POWER for extended operation has been achieved. .02 07 A i

CPSESMarA-up ofCTS 3N.2 3N 2-6 7/19/98

CHANGE NUMBER RSEC DESCRIPTION consequences of severe transients with unanalyzed power distributions. The ccmpletion time of 72 hours is sufficient considering the amount of wark required to be done to reduce the setpoints, the small likelihood of a severe transient in this time period, and the prompt reduction in THERMAL POWER required upon discovery of the out of limit condition. 02 03 M The required ACTIONS are re written for consistency with NUREG 1431. The specific changes include the addition of a requirement to be in at least MODE 2 within 6 hours should any of the ACTIONS not be completed within the required time period. This requirement is more restrictive than the previous requirement to enter LCO 3.0.3, which allowed I hour before the 6 hour shutdown requirement became effective. 02 04 M Consistent with NUREG 1431, F/(Z) must be verified to be within limits whenever Fa(Z) is measured, not just at the time of target flux determination as required by the current TS. Hence, this change imposes requirements which are more restrictive than the current TS. 02 05 M Consistent with NUREG-1431, F/(Z) and F/(Z) must be verified to be within limits prior to exceeding 75% RTP after each refueling. This requirement is not explicit in the current TS. The TS are made more restrictive by stating this requirement. 02 06 AM Consistent with the bases of current TS 4.^.2. which 0 3.2 3 allows 24 hours for completing surveillance rcquirc: cats l that become opplicMe whcn on exception to Specification 4.0.4 is clicacd, the frcquency for assessing Fa(Z) is clarificd by requiring that thc ac;surc cat be peefor;cd within 24 hours after reaching equilibrium conditions. Inithe;ITS;SR 3:2.1;11andlSR;3.2l1:2,la time limit;fotassessing:Fa(Z) after;rpaching; equilibrium conditions;istspecified;2Becaugthe CTS;does;notLhave1such altimeRestrictionEthisLchangelis;morejrestrictiveOThe. time linj ttforicomp]etioniof ~ thi sisurveill anceM 24l hours 2 following thelestablishment;of 1equilibrium 1 conditions;Ehas;been selected'basedion; plant;experienceCTwenty four hoursiisia reasonableitime; for [obtaini ng :and: eval uati ng ' a ;fl uxJ map;and I thenicompleting;the^ required procedural? steps; associated;with this: surveillance;2 [urther;';;the:24"houritimeilimit does;not allow;for:plantioperationiinian; uncertain condition for'a protracted time period! The: time limit'.ofl24; hours 11s consistent _withl Amendment 1NoJ116;for WolfiCreekjinLwhichsthe NRC approved l allowing the performance;ofJa? flux map;24; hours afterachieving(equilibrium; conditions? from)a1 Thermal Power reductionirequired.withiQP,TR; determined to; exceed liO2; CPSES Description of Changes to CTS 3N.2 4 July 29,1998

CHANGE , NUMBER RSBC DESCRIPTION acceptable becausea F is not necessarily outside of its limits and the probability of an event during the 4 hour period is low. The removal of the requirement to reset the AFD alarm setpoints reflects the deletion of a requirement fecm the Technical Specifications. The means of monitoring compliance with more restrictive AFD limits is reflected in the appropriate operating procedures and is beyond the level i of detail required of Technical Specifications. 02 09 M Not applicable. See conversion comparison table (enclosure l 3B). 02 10 Not Used. 02-11 LG Not applicable. See conversion compaaison table (enclosure 3B). 02 12 A Not applicable. T.se conversion comparison table (enclosure 38). 02 13 LC Not used epplicabic. Sec conversion cc,T,parison table

                                                   'ancic;urc 30).                                                   0 3.2 3 02 14               M          Not applicable. See conversion comparison table (enclosure 38).

03 01 LG Moves the details of the F",, limits to the COLR. Previously, the equation for the dependence of F",, on THERMAL POWER had been located in the LC0 and the COLR. The full power limit value of F",, and the power factor nultiplier had been located only in the COLR. 'Now, the equation is also located only in the COLR. Definitions and details of the measurement. including the treatment of uncertainties, are moved to the BASES. The RE0'JIRED ACTIONS are re written for consistency with NUREG 1431. The changes are acceptable because they remove details not required to be in TS to support operational safety. 03 02 LS 5 Revises the completion times to be consistent with NUREG 1431. The adequacy of these completion times are discussed in the applicable BASES section of NUREG 1431. In summary. 4 hours (vs. 2 hours in the current TS) is provided to attempt to restore F", to within its limit or to reduce power to below 50% RTP. l 03-03 M The Requirement to reduce power to less than or equal to 5% ( RTP (exit Mode 1) within the next 6 hours is added in lieu of i the use of LC0 3.0.3. This requirement is more restrictive ( than the previous requirement to enter LC0 3.0.3, because LC0 3.0.3 allowed I hour before the 6 hour shutdown requirement CPSES Description of Changes to CTS 3N.2 S July c.1998

A e C ec L R n tn i - i L s s s

                                                              -                        + ;

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I CHANGE NUMBE8 JUSTIFICATION Condition D, the breakpoints for the applicability of the surveillance in the notes in improved TS SR 3.2.4.1 and SR 3.2.4.2 are modified to be applicable at less than or equal to 75% RTP, and greater than 75% RTP, respectively. This is an administrative change that retains current TS requirements. 3.2 10 Consistent with TSTF 110, this change moves requirements for increased surveillance frequencies in the event of inoperable alarms to licensee controlled documents. This change is acceptable because it removes requirements regarding alarms and alarm responses that are not necessary to be in the TS to protect public health and safety. 3.2 11 This change, applicable to LC0 3.2.3, Axial Flux Difference (Constant Axial Offset Control plants only), collects 3 LC0 Notes and one Applicability Note into

• Notes" list under the LCO. The revised presentation enhances clarity and usability. The Applicability Note is inappropriately located since it takes exception to the LC0 requirement. The Note is moved to the LC0 Notes. This change is consistent with traveler TSTF 164.

3.2-12 t with current Technical Spccif tcations. The 0 3.2 3 required time for completion of a flux map for determination of the heat flux hot channel factor is changed from 12 hours to 24 hours after achieving equilibrium conditions. The proposed change affects SR 3.2.1.1 and SR 3.2.2.13J2T1;2. Based,on plant experiences the proposed time (24 hours) is a reasonable time period for obtaining and; evaluating a;fluximap and then. completing the:procedurallstep.s1 associated .withithis1 surveillance; iFurther; theT24 hour; time; period;doesinot ' allow:for plantioperatjoniinf an uncertainJconditionjfor;afprotractedl time, period 7 thc compietion of thc surveillance and docs not allow for plant op ratica in an uncertain condition for a protracted tiac period. This change is consistent with the the Tcchnical Specificatica rcquirc; cats of speenication 3.0.4 (and associated Sasc;) that alicw 24 haurs fc- ! th; completica of a surveillance ^ after picr ^quisite plant conditions arc atte ncd and for ahich ar, exccption to ciflcatica 4.0.4 was provided.

                                            ~.

3.2 13 This change retains the CTS for the performance of peaking factor determinations following plant shutdowns. The CTS through the exemption to specification 4.0.4, allows prerequisite plant conditions to be obtained prior to requiring that the 0 3.2-4 surveillance be completed. The note wasiincorporatedsto address;thelrare[ situation wherel1during1aLmid cycleishutdownl l throughjfurtherfreview;ofothe; previous; surveillance,11tlwas l CPSES Differencesfrom NUREG-1431 - ITS 3.2 3 7/29/98

CHANGE NUMBER JUSTIFICATION 3.2 15 Thh change Worporates industry traveler TSTF 109. Action A.2 would requ' e QPTR be determined rather than performing a specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1. Theinote!andfrequency for[ sri 3;2;1 2iarefevised 0 3.2 10 tolbe.LconsistentMth;typica][ presentation;formatsithat providolforia; perjodfofitieelaft;erlestablj shing.[ conditions? The note for S"o^-G.4.2 is che.n,;d to require perferna;; if en; "er ;;;;re" OPEnputa era ineperebla. These changes are acceptable because they clarify the ISTS regarding frequency and use of incore flux monitoring for QPTR measurement. The changes reflect tl.at incore detectors provide an acceptable QPTR determination during all plant conditions. 3.2 16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that Fa(Z) is within its limit. Fa(Z) is approximated by FE(Z) (which is obtained via SR 3.2.1.1) and Fs(Z) (which is obtained via SR 3.2.1.2). Thus, both Fj(Z) and F#(Z) must be established to verify Fa(Z), This change is consistent with J traveler WOG 105.

                                                                -                                               w 3.2 17                            Not used, applic:ble t; SPSSS. S;; ;;nversion ce.T,pariscr'       0 3.2 3 table (;nclosure S").

3.2 18 Not app e-CPSES. Saa conversion n-perison teMe (enclosure 6B). 3.2 19 Not applicable to CPSES. See conversion comparison table L (enclosure 6B). 3;2f20 Thisichange;jncorporatesiTSTFell2 (ire;Edeletesithe; Note:in o.3.2 1 l Condition;D_Which_ required;thatirequired action l D;11be coepleted;Whenever1ConditjonIDI1sientered)lCIDeletjng;the Notepermits[exitingsthelrequired_ action;when;poweriisreduced below1500RTP C ExitingitheTrequiredlactionjis; appropriate:because operationibelow;50t'RTPJ1sinoticonstrained;with1 respect 1to AFD] CPSES Differencesfrom NUREG-1431 - ITS 3.2 4 7/1988

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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.2-4 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.2.2 Nuclear Enthalpy Rise Hote Channel Factor CTS 3/4.2.3 Nuclear Enthalpy Rise Hot Channel (All FLOG Plants) DOC 02-07-A JFD 3.2-13 SR 3.2.2.1 NOTE and related Bases. Comment: Justify the need for the note related to permitting power ascension after shutdown to a level at which a power distribution map is obtained, it appears that this note is unnecessary, considering the phraseology of the SR Frequency ("Once after each refueling prior Thermal Power exceeding 75% RTP"). Explain the need for this note. The SR 3.2.2.1 Bases also mentions

 "(leaving Mode 1)" which appears to be the incorrect mode.

FLD's Response: The note, as described in JFD 3.213, was incorporated to address the rare situation where, during a mid-cycle shutdown, through further review of the previous surveillance, it was determined that the surveillance was invalid. The amended Note would be required to retum the reactor to a power level at which a new surveillance could be performed. The " leaving MODE 1" clarification is based on the Applicability of the LCO (MODE 1, only) and is intended to avoid confusion in a scenario where the plant may be taken off-line (typically, MODE 2), but not " shutdown" (commonly considered to be MODE 3 or lower). ATTACHED PAGES: Enci 6A 3 & 3a ' i l I l l - ---- _ ---- -- - - - 1

CHANGE NUMBER JUSTIFICATION Condition D, the breakpoints for the applicability of the surveillance in the notes in improved TS SR 3.2.4.1 and SR 3.2.4.2 are modified to be applicable at less than or equal to 75% RTP, and greater than 75% RTP, respectively. This is an administrative change that retains current TS requirements. 3.2-10 Consistent with TSTF-110, this change moves requirements for increased surveillance frequencies in the event of inoperable alarms to licensee controlled documents. This change is acceptable because it removes requirements regarding alarms and alarm responses that are not necessary to be in the TS to protect public health and safety. 3.2 11 This change, applicable to LCO 3.2.3, Axial Flux Difference (Constant Axial Offset Control plants only), collects 3 LC0 Notes and one Applicability Note into " Notes" list under the LCO. The revised presentation enhances clarity and usability. The Applicability Note is inappropriately located since it takes exception to the LC0 requirement. The Note is moved to the LC0 Notes. This change is consistent with traveler TSTF 164. 3.2-12 Consistent with currcat Technical Specificaticas, The 0 3.2 3 - - - - required time for completion of a flux map for determination of the heat flux hot channel factor is changed from M hours to 24 hours after achieving equilibrium conditions. The proposed change affects SR 3.2.1.1 and SR 3.2.2.13.~2.1;2. Based lon plant experience l the proposed time (24 hours) is a reasonable time period for obtaining;and;evaluatingla1 flux;mapiand thenl comp 1etjng the: procedural stepsLassociatedfyith1thisisurve111ance.( Murther l the'24ihour, time,periodidoes;not;allowifortplantioperationiin;an i uncertainrconditioniforial protracted time; period; the cocpletion of the surveillance and docs not allow for plant operatica in an uncertain condition for a protracted time period. This chant,c is ccasistent with the the Tcchnical Specificatica requircccats of specification 3.0.4 (and associated Bas s) that allow 24 hours for th; compictica of a surveillance after prerequisite plant conditicas arc attained and for which an exception to specification 4.0.4 was providcd. l 3.2 13 This change retains the CTS for the performance of peaking factor determinations following plant shutdowns. The CTS through the exemption to specification 4.0.4, allows prerequisite plant -- conditions to be obtained prior to requiring that + 0 3.2-4 i surveillance M Wo+aufhej note _was; incorporated;to address 1the rare; situation. where;;duringla.mid cycleishutdown; throughtfurther; review;ofithefprevious1 surveillance 1 1.t was l 3 7/29/98 CPSES Differencesfrom NUREG-1431 - ITS 3.2

CHANGE NUMBER JUSTIFICATION termined?that the? surveillance;was'" invalid.' The1 mended l Notewould[behequired;toireturnithelreactorito:a: power;;1evel l 0 3.2 4 at which ainew; surveillance?could:belperformed? 3.2 14 Not applicable to CPSES. See conversion comparison table (enclosure 68). l l l l I l l l 1 l I l i l 3a 7/19/98 CPSES Differencesfrom NUREG-1431 - ITS 3.2 L

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.2-6 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.2.4 Quadrant Power Tilt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (All FLOG Plants) DOC 04-01-A JFD 3.2-05 ITS Required Action A.5 Comment: The ITS proposes to change the STS wording for Required Action A.5 from

                      " Calibrate excore detectors to show zero OPTR," to " Normalize excore detectors to eliminate tilt," based upon WOG-95 (and rejected TSTF-25). A preferred wording would be that proposed in the Comanche Peak CTS mark-up, " Calibrate excore detectors to show zero Quadrant Power Tilt." What is status of WOG-957 FLOG Response:                                                                                                                                               Traveler WOG-95 was transmitted to the NRC in February 1998 as TSTF-241. The FLOG is incorporating TSTF-241 including the latest revisions discussed ct the June 1998 WOG MERITS Mini-Group meeting. These revisions corrected errors made during the development of TSTF-241.
                                                                                                                                                                                                                                                         -l Additionally, Wolf Creek submitted a License Amendment Request to CTS 3/4.2.4, Quadrant
                                                                                                                                                                                                                                                        ]

Power Tilt Ratio, on February 4,1998 which was approved on April 27,1998 in Amendment No. l 116. This amendment incorporated the changes proposed in TSTF-241. The FLOG believes that it is appropriate to incorporate the proposed TSTF-241 changes based on the NRC approval of the Wolf Creek amendment request. ATTACHED PAGES: Encl. 2 2-10 Enct.3A 7, 8 I Encl.3B 5, 6 Encl. 4 1 l Encl.5A Traveler page,3.2-14, 3.2-15 Encl. 5B B 3.2-33, B 3.2-33a, B3.2-34, B3.2-35 Encl.6A 1, 1a, 4 Encl.6B 1, 3 l l I l 4

POWER DISTRIBUTION LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: H0DE 1, above 50% of RATED THERMAL POWER *. i

01-01 A- l ACTION: I

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02: 1 1

1. a) Within2hourdereachQPTR' determination'5Teduce t 04 01 A THERMAL POWER by at least M trom KAltU IHtKMAL POWER for 0 3.2 6 each it of QUADRANT POWER TILT RATIO in excess of 1 I l

b) At least once per 12 hours calculata the OUADRANT POWER l TILT sm 04 01 A

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                                                                                       . e n-    .a s.    ,  and                                                      'I c)       Within 9Ah^'rSMter, achieving.,equilibriumcondito                                             j (Trom 'a1THERMALc POWER " reduction wr Action a ,1. a) ,

once 04 01.A per 7 days thereafter, confirm that the Heat tlux Hot 0 3.2 6 Channel Factor Fa(Z), is within its limit by performing I Surveillance Requirement 4.2.2.2 and confirm that Nuclear Enthalpy Rise Hot Channel Factor, Fln, is within its limit by performing Surveillance Requirement 4.2.3.2.;

2. Prior to increasing THERMAL POWER above the limit of Action a.1:

a) Re evaluate the safety analyses and confirm that the results remain valid for the duration of operation under this condition, and then b) dlormalize Cclibr 'D xcore detectors to show cero OPE 0A. 01 A

                                                                        @cc; c;;t ;-. ,;jlt res. tore QPTR, withinM-          -

o.3.2 6

3. ter A.cti= t? ic enmn19 tad and withinMafter ~achiev ilibrium" conditions;at'RTP_not;to exceejd m as ci rcacning

                                                               "^Trn m := r ,;d, J,,ith'- A9 burs 01 afte^ inaeasing lHERMAL POWER above the limit of ACTf0N a.1, confirm that a ) is within its limit by performing Surveillance Requirement 4.2.2.2 and that Flu is within its limit by performing Surveillance Requirement 4.2.3.2**; and

4. If the requirements of a.1, a.2 or a.3 above are not met, reduce THERMAL POWER to 5 50% of RATED THERMAL POWER within the next 4 hours.

Scc Special Test Exceptions Specification 3.10.2. -01 01-A5

                                               **      Action . a;3; ;shall' be, completed; whenevert Action; a;2.b) Lis, performed.

04 01 A 0-3.2 6 l l CPSESMark-up ofCTS 3N.2 3N 2-10 7/29/98 L_- - - - - - - - - - - - - - - - - - _ _ _ _ _ _

CHAroE NUMBER HSjiG DESCRIPTION 03 10 LG Not applicable. See conversion comparison table (enclosure 3B). 04 01 A Clarifies that when the excore detectors are calibrated, theQuadrantPowerTiltisrestoreditoiwithinjimit 03.24 nrod out. (The QPTR is normalized to unity.) This requirement from NUREG 1431 as moaified by TSTF 25241, is consistent with the current TS ACTION requirements for , verifying QPTR is within limit during power escalation I subsequent to identifying and correcting the cause of OPTR out of limit. Additionally 3he; actions [are modified to ( Ptoy1delthefappropriate'allowanceJfonesubsequent power j reductionsLbased;on;subsequentJdetermination;ofiQPTR_.i 04 02 LS 10 Not applicable. See conversion comparison table (enclosure 38). 04 03 LG Not applicable. See conversion comparison table (enclosure 38). 04 04 LS 12 The current TS does not contain any provisions for determining QPTR with more than one inoperable input; thus, LC0 3.0.3 would be entered and the plant would be shut down. The proposed change would allow for the use of the movable incore detector system to determine an equivalent QPTR with one or more inoperable excore detector inputs to the QPTR calculation. In addition, the frequency is clarified by a note which says that the SR is not required until 12 hours after input from one or more Power Range Channels is lost. l If the movable incore detector system is used to determine an l equivalent QPTR, the QPTR calculation is not based on information gained from any operable excore indications and, therefore, is independent of the number of operable excore detectors. The frequency specified in the current Technical Specification for the determination of an equivalent QPTR

with the movable incore detectors (every 12 hours) would be retained. Further justification for this Frequency is based on the fact that under normal circumstances, QPTR would not be expected to change significantly within a 12 hour period.

If a significant change in QPTR were to occur, it would l likely be the result of control rod misalignment which would be detectable immediately by means of the rod deviation l monitor or rod bottom lights. 04 05 LS 11 Not applicable. See conversion comparison table (enclosure 3B). 04-06 LS 13 Not applicable. See conversion comparison table (enclosure 3B). CPSESDescription of Changes to CTS 3M.2 7 July 29,1998

j CHANGE NUMBER H2iG DESCRIPTION 04 07 A Not applicable. See conversion comparison table (enclosure 38). i i 04-08 Not used, i i 1 04 09 A Not applicable. See conversion comparison table (enclosure 3B). 04 10 LS 14 Not used Copplicabic. Sec conversica comparison table (cnclosurc 30). 0-3.24 l r 04111 A Not:applj cabl e.mSeeiconver.sion; comparisonitable 3 enclosure 3B)M 4 05 01 LG The designation of how instrument uncertainties are treated i (nominal, in the analysis, or in the development of the TS j limit) is moved to the Bases. The movement of this level of detail out of the specification is consistent with NUREG-1431 f and is an example of removing unnecessary details from the TS in accordance with 10 CFR 50.36. 05 02 LS 7 The requirement to verify that the total RCS flow is within limits using the plant computer or elbow tap output voltage on a monthly basis is deleted. This action was included in the TS based on the potential drift of the RCS flow indication. The essence of this activity is performed on a quarterly basis through the Channel Operability Test performed in accordance with the Reactor Trip System surveillance. , i 05-03 LG Consistent with NUREC 1431, thc requirc;;at to CP 3.2-001 perfor: a C"#NEL CALIBRATION ca the RCS ficw acters  ; et icast once per 18 month, and The requirement to j normalize the RCSiloop;flowfate indicator _s2channcis arc is j moved to the Bases for the RCS flow low reactor trip 1 functioninITSSection[3.341]. I 05-04 LG Consistent with industry traveler TSTF-105, the explicit ] requirements that the RCS flow be measured through the use of j a precision heat balance measurement and that the l instrumentation used in the performance of the calorimetric l flow measurement be calibrated within a specified time period of performing the measurement is moved to a licensee controlled document. The requirement to verify that the RCS i flow is within limits remains within the Technical j Specification. This is an example of removing unnecessary { details from the TS and is acceptable based on the guidance 1 l provided in 10 CFR 50.36. 4 l l CPSES Description of Changes to CTS 3N.2 8 July 29,1998 t [ ___ _ ___________________ _ - _- -__ -

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t l NO SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) CONTENTS I I. Organization ........................................ 2 II. Description of NSHC Evaluations. . . . . . . . . . . . . . . . . . . . . . 3 III. Generic No Significant Hazards Considerations "A" Admini strative Changes . . . . . . . . . . . . . . . . . . . . . . . . . 5 "R" - Relocated Technical Specifications. . . . . . . . . . . . . 7 "LG" Less Restrictive (Moving Information Out of the Technical Specifications) . . . . . . . . . . . . . . . . 10 "M" - More Restrictive Requirements. . . . . . . . . . . . . . . . . 12 IV. Specific No Significant Hazards Considerations "LS" LS 1................................................ 15 LS 2................................................ 17 LS 3................................................ 20 LS 4................................................ 23 LS-5................................................ 26 LS 6................................................ 28 LS 7................................................ 30 LS 8............. .................................. 32 LS - 9. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS - 10. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS 11. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS-12............................................... 35 LS 13. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not A plicabl e LS 14....................................No usedAppi k able LS 15. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e r i CPSES No Significant Ha:ards Considerations - CTS 3N.2 1 7/29/98

i ' INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.2 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-24 Not Incorporated NA Not NRC approved as of traveler cut off date. TSTF 95 Incorporated 3.2 06 Approved by NRC TSTF 97 Incorporated 3.2-07 Approved by NRC TSTF 98, Rev. ;1 . Incorporated 3.2 03 TSTF 99 Incorporated 3.2 08 Approved by NRC TSTF-109 Incorporated 3.2 15 Approved by NRC TSTF 110 Rev. 2;-1 Incorporated 3.2-10 Approved ~by NRC u; .2 op I numasummmmed TSTF 112, Rev. 1 Net Incorporated NA 3.2 20 N^t NRC approved 2-1 as of t~eveke un's cut-off date. I Approved by;NRC;; Applicable to'CAOC p1antst (CPSES'only) TSTF-136 Incorporated NA Approved by.NRC l R.: .2 0M TSTF 164 Incorporated 3.2-11 Applicable to CAOC only. (CPSES) _ WOC-Z. proposed Incorporated 3.2-05 Rcv. 2 TSTF-241 3.2469 0 3.26 p l ) ReVE1 _ __ W0G 105 Incorporated 3.2-16 l 1 I , I l t_____________________________________ . - . . _ _ . _ _ _ _ _ _ _ . - _ _

QPTR 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR) LCO 3.2.4 The QPTR shall be s 1.02. APPLICABILITY: MODE 1 with THERHAL PCWER > 50% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

                                                                                                                              ~     ~

A. QPTR not within limit. A.1 Reduce THERMAL POWER 2h r .fter 2 3% from RTP for each .achiQPTR 3.2 05 1% of QPTR > 1.00. erminatio 0-3.2-6 E A.2 "erfor, SR 3.2.4.1 Once per L3.2-153 Determine QPTR 12 hours - i suum ....-.-n -. 3.2 05 Si fre;;; "T" for CCCh II 0-3.2-6 OfQ"" =1.00. ) m A.3 Perform SR 3.2.1.1; SR13;;2 % 2 and 24 ho c<.ieving k """" 3.2 05

                                                                                                                        -   0-3.2-6 SR 3.2.2.1.                 equilibrium conditions

{ from a; THERMAL POWER E9 reduction" pet 0 3.2-6 A m - g c.3.2-16' Once per 7 days thereafter M A.4 Reevaluate safety analyses and confirm Prior to increasing results remain valid THERMAL POWER above for duration of the limit of operation under this Required r condition. Action Al/end ' E9  !

                                                                                                 /Fr2 g                                                -

0-3.2-6 l (continued)  ! CPSESMark-up of NUREG-1431-ITS3.2 3.2-14 7/19/98

QPTR 3.2.4 CONDITION REQUIRED ACTION COMPLETION TIME l A. (continued) A. 5 - l -/h.erfornRequ

                                                                    - - - NOTh - - i re Wtion A.5 only after Required Action A.4 is I             completed.                                             ,
                                                                      ._ Required action A.6 shall;be ; completed WheneverJRequ1_ red
                                                                                                                           /

Act. ion A'.511s 3.2 05 rformedj 0-3.2-6 valibrate execre Prior to increasing detectors to show Icr. THERMAL POWER above OPTR Normalize;.excore \ the limit of I detectorsito eli;;;in;te Required t4Mr restore.<QPTR;to Action A. ~ and l withinflimitt rfG E9 s 0-3.2 6 6NQ A.6 No - - Perferit t erfor  % R ed Action A.6

                                                                      ;;st b; c.,7.pleted h;3                                                                                     [3.2-6    '

caly aftcQnly:afterb Required Action A.S is (foir.pleted completed. jT.p;;r;r,ted I Perform SR 3.2.1.1; Within 24 hours -3.2-16v SR::3.2.1;2 and paffacchich SR 3.2.2.1. (achieving I equilibrium conditions _at RTPlLnot.to exceed 3.2 05 QB 0-3.2-6 k 48 hours I after increasing THERMAL POWER above ! the limit of Required Acti A.1 60

                                                                                                                                 ; r.d A.2 e                                      0-3.2-6 CPSESMark-up of NUREG-1431-ITS3.2                                  3.2-15                                                                                                  7/29/93

QPTR B 3.2.4

     -BASES ACTIONS           Ad With the QPTR exceeding its limit, a power level reduction of 3% RTP for each it by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours allows sufficient time to identify the cause and correct the tilt. Note that the power reduction itself may cause a chang in_the tilted conditioL./The;maximunjallowabletThtiemL;PUWER evellinitial]y;determinedibymRequired; Action:A;1*may;be                                                                                 0 3.2 6 affactedihylsubsequent, determinationslof;QPJR{S Increases;j n QPJRLwould[ require?a3NERMAL; POWER                                                                i   reduction;within 2; hours;of QPTRTdeterminationRifanecessaryito; comply with;the: decreased:naximum a110wablelTIERNAL_POWEREleve120ecreasesling]Wwould allow! raising thelmaxieus;a]1owable3NERNAL. POWER!1evelsandlincreasing; THERMAL POWERiupitoithis;revjsedt]imiti w                         m.

After completion of Required Action A.1, the QPTR aler;;; ;;;;y :till b; in it eier;;;;d state;mayf.still: exceed;;1ts!)imits. A such, any additional changes in the QPTR are detected v ~'94 a check of the OPTR once per 12 hours thereaft -If th oat" C atinues to incr;;;;, E'""i "0WER hq;L" M redu;;d Qc;;rdini;1M12 hour Completion Time is sufficientf because any aaanlonal change in QPTR would be relatively slow. T peakingfactorsFin and Fo(Z) are of primary importance in ensuring that the power distribution remains consistent with the initial conditions used in the safety analyses. Performing SRs on F1, and Fa(Z) within the Completion Time of 24 hours after3chieving equil ibrium: conditionsifrom:alTtlEkMAL~ POWER; reduction lper RequiredfActionTA11 ensures that these primary indicators of power distribution are within their respective limits. l03.26 1 Equ111brjumiconditionslarelachieved;when the:coreiis sufficientlyistablej atithelintended; operating 1conditjonsito support;Guxleapping; A Completion Time of 24 hours afterJachieving i equilibrium; conditions;frosJaHHERMAltPOWERreduction?per. Required l Action X I takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perform a flux map. If these peaking factors are not within their limits, the Required Actions of these Surveillance provide an W 1nued) CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-33 7/29/98

                                                                                                                                                             -_____-__A

QPTR

                                                                                                                           )

B 3.2.4 i BASES 1 l 1

appropriate response for the abnormal condition. If the QPTR remains l above its specified limit. the peaking factor surveillance are required each 7 days thereafter to evaluate Fin and Fa(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the OPTR limit.

i f i l 1 l (continued) CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-33a 7/1988 l L______-_____-______-__-____- - _

i. t QPTR B 3.2.4 i BASES ACTIONS M l (continued) l Although FL and Fa(Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the i validity of the safety analysis. A change in the power distribution l can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit it i does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power ! distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure l that, before increasin L POWER to above the limits of Required Actions A. ;. .O . the reactor core conditions are consistent with the ass ons in the safety analyses. 0 3.2 6 c M -

                                                                                             )

l If the QPTR bcs cccded remainslabovelthe 1.02 limit and a j re evaluation of the safety analysis is completed and shows l that safety requirements are mets the excore detectors are _ recalibrates to d : =; ^"" normalized"tofestore;QPTRit M H Ehin limit ~ h y lor to increa g THE o mw rn Je the limit of Required Actions A- and

m. Normal izationjj s;accomp p .t-a.m.n..a; sannet that; t

                        .i.n dicated;QRTRifollowjng;normaljzatioEis;near:1600g               1s done to etect any subsequent significant changes in QP m

Required Action A.5 is modified by a two' te; states thattthe;excore_ detectors are rea _ that ^ l, y crestore to restore urwto within limits . .; g-@c ^"T is r,ct ;;rced cut until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within_the safety anal sis acen=ntions (i.e., Required Action A.4)foteJ21 state at21f; Required;ActionJ6;61.s;perj'ormedEthen_ s Required: Action Aq shall(be:performedERequiredlActjon A16[normaljzes;.the;excore detectors 1toirestore10PT[to[within21mitRWhichirestoresicompliance wjth:LCOI3;;2:4;?;Thus',1 Note:2 prevents; exiting;the Actions priorfto completingiflux:sappingito; verify _peakingifactors;petLRequired Action

                         $?72 This Note is Theselnotes:;arelintended to prevent any ambiguity the required sequence of actions.

(continued) CPSESMark-up of NUREG-H31 Bases-ITS3.2 B 3.2-34 7D9/98

QPTR B 3.2.4 i BASES l l Lfi Once thejeXcore;; detectors;are:normalizeditoirestore;QPTRito;within limit cli; int;;thq;i-dicctix;cuj]t flux tilt is ccrced cut (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution et-RTP is consistent with the safety analysis assumptions, Required Action A.6 requires verification that Fo(Z) and F1 are within their specified limits within 24 hours of l (achieving equilibrium conditionsf8t Mir, reccning lU N As an_ added_precautiodif[thehcAre! power [doesinotreachiequilibrium 0 3.2 6 _ itions2at3TP;within 20 hours;but;isdncreased: slowly; then1the; peaking:factorisurveillancesimust;betperformedNithin 48f hourslafter71ncreasing.THERMALT P0ER"ahave the climit~of Reau tiontA~;1' . tF.c arc powcr Occs not reach P,TP within 24 hours but is incr - cd slowly, then the pa king factor survcillancc; must be perforgd w"- 9 ' = e ' += "= wha the ascent to =- was

                                    . E ttA pChinij Zt0i N rj U Gtie. GR J.; G c F forF~d Wit A 24I.c;-^ dKt; f.c;Acilib j;.GER.?cadititj9i.;,1=Mic;' ti; 40fcut; i; cilH ' r the n viction ^fith; ycijficti=;

Thi ompletion Times-eee is intenaea to allow auequale -inue t j e e THERMAL POWER to above the limits of Required Actio .1 6 .im A. while not permitting the core to remain with unconfirmed er distributions for extended periods of time. Required Action A.6 is modified by a Note that states that the 0 3.2 6 l peaking factor surveillance m y caly bc donc after mus.t be l completediwhenjthe excore detectors have been calibrated to t sh= cro tilt normalizeddo restore QPRto within limit- ! h indi sted.t g e., Required Action A.5). The intent of this Note is to nave the peaking factor surveillance performed at operating power levels, which can only be accomplished after the excore detectors are normalized;colibrated to restore;QPTR'to;within limit. show ccro tilt =d the a rc returned to powc-L1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or l condition in which the requirements do not apply. To achieve this status THERHAL POWER must be reduced to < 50% RTP within 4 hours.  ! The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems. (continued) CPSESMark-up of NUREG-1431 Bases-ITS3.2 B 3.2-35 7/29/98

JUSTIFICATIONS FOR DIFFERENCES FROM NUREG 1431 i NUREG 1431 Section 3.2 This enclosure contains a brief discussion / justification for each marked-up technical change to NUREG 1431, Revision 1, to make them plant specific or to incorporate generic changes resulting from the Industry /NRC generic change process. The change numbers are referenced directly from the NUREG 1431 mark ups. For Enclosures 3A, 3B, 4, 6A, and 68, text in brackets "[ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcommittee (JLS) 1 plants. Empty brackets indicate that other JLS plants may have plant specific l information in that location. CHANGE NUMBER JUSTIFICATION 3.2 01 Consistent with the current CPSES TS, substitute the overpower N- I 16 for the overpower AT reactor trip function. Overpower N 16 performs the equivalent function for CPSES which does not have overpower AT. 3.2 02 Consistent with the current TS, retain the requirement for , i performing Fa after a 20% change in power (vs. the 10% value i specified in the ISTS). 3.2 03 Consistent with TSTF 98, Rev. 1., the factor by which the Fa must be adjusted on increasing Fa measurements is moved to the COLR. This change is acceptable because the factor is normally contained in the COLR, and it removes detail not required to be contained in TS. 3.2 04 SR 3.2.3.2 is applicable to plants using the Westinghouse l Constant Axial Offset Control methodology (CPSES). TU Electric , uses the methodology described in RXE-90 006 P A, described in current TS 6.9.1.6b. With this methodology, verification of that the peaking factors are within limits is required to be performed at the time that the target flux difference is determined. Therefore, it is appropriate to tie the target flux difference surveillance frequency to the frequency at which the Ff(Z) peaking factor is verified. i (([I5I?35f_5I!!!II)Ib,bN!bI'II5_ !5 !I,*. E C. . . . . .J^'.".L:,

                                                  ",'.:".,'.. ".i", ,..". L ". ".1. ,"'.:r. =...,_",. ",, ~. , ' .'~"1. , :,'_' ., ~': ' . .: '
                           ;;;;ptebl;.

Cons 1stentyj th;TSTFj241XISTSJ3;2;4 g 0uadrant fewerlT11 tiRatio t istrevisod to; i provide.more! appropriate! Actions G Required; Action A;21containsla; redundant;actjonito, reduce 7 5 C POWER E Thjs CPSES Differencesfrom NUREG-1431 - ITS 3.2 1 7/29/98 _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ~

CHANGE NUMBER JUSTIFICATION redundant; action;is.1 deleted and;theiTHERNAL POWERilimitiot 0 3.2 6 RequiredLAction;A;111sirevised;tofprovidelthe; appropriate allowance!forisubsequent. power reductions; based;onisubsequent determination;of R TR C [] The; Completion 1 Time;ofmRequired, Action A;3: requires 1the.; peat' a factors;to ;bej verified _within

• 24: hour.s;ofi schievingiequ.111brium conditions;with1TERMAL POWER reduced;byiRequired ActiotEA.1^.;?.,In the:currentLRequired; Action gal;1gnificantifractionfofithe 24

                         .hourscould1bel; spent;waitingiforlthe;plantito) stabilize:at;'the;new powerDevelgeavinginsufficientitime;tolaeasureLandLanalyzelthe peaking; factors (orLresulting;inithe peaking factors;being_. measured when,the plantjis;enotistable; yielding 11naccurate71nformation.,

Since'the;peakingifactorsare;ofLprise;1mportancenthe; proposed change s williallowlsu_fficientitisedtoiobtainlan1 accurate measurement l((T] Required: Action:[A.5]lis^revisedito;addfalnew; Note; stating [ Required: Action![A;6]isha_ll;;Mcompletedjif(Required Action,[A;5] is(performedGAs:di.scussed;inlSectionII:.3:ofitheJITSRan Actions Condition l remains:inieffactLand;theiRequiredjActionsLapply!until, the: Condition no' longer;exitsiorithelunitrisinotlwithin:the:LCO Appl icabilitytTherefore n whe;11 Required: Action '[AJ 5]11 s completedRQPTRishould:be;back withinilimitiand;the Lp0:may;be exited 7EAdding this! Note: ensures;.that;theLpeaking;fac1. ors;are verified after" normalization 1oft thelexcore;detectorsh Additionally;; Required; Action;[A.5]iis: revised;to; state " Normal.1ze excoreidetectors itoirestore . QP1R Ltoiwithin il imitC:. Normal ization 1sf accomplished 11nisuchlasmannersthat!the;indi.cated QPTR;;is:near 1;00MThusRthelabsencelof(altiit willimanifestfitselfas;QPTRi= liO0fratherithan:;zero;since1quadrantipoweritiltiislexpressedsas a ratio l.C Alsonfromia;] iteral"compliancelstandpointMtheitilt l cannot;be restoreditolexactly:1;00l1[f] 3.2-06 Cmsistent with TSTF 95, the time allowed for resetting the power j range neutron flux high setpoint if Fa or F % is outside their limits is extended from 8 hours to 72 hours. As written. CPSESDifferencesfrom NUREG-1431 - 11S3.2 la 7/2988

CHANGE NUMBER JUSTIFICATION 3.2 15 This change incorporates industry traveler TSTF 109. Action " ? would require the QPTR be determined rather than 0 2.2 10 perfo % .ig a specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1. Jhelnote;and;Frequencyffor;SR:312;.422;are reyisoditolbe;consistentIMth;typicallpresentationiformatsithat providolforla:petiodtof;tjuerafteelestablishjngiconditions; He note-for SR 3.2.4.2 is chenged to requira perferr.ence if one "er ore" OP" input; ere ineperebla. These changes are acceptable becease they clarify the ISTS regarding frequency and use of incore flux monitoring for QPTR measurement. The changes reflect that incore detectors provide an acceptable QPTR determination during all plant conditions. 3.2 16 This change would require that bcth transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Pequired Actions is to verify that Fa(Z) is within its limit. Fe(Z) is approximated by Fj(Z) (which is.obtained via SR 3.2.1.1) and F#(Z) (which is obtained via SR 3.2.1.2). Thus, both Fj(Z) and F#(Z) must be established to verify Fa(Z). This change is consistent with traveler WOG 105. 3.2 17 Not used. pplicabic t; CPSSS. Sc ;;nversion ca.p ri: a 0 3.2 3 table (;r.cicsurc SS)- _ 3.2 18 Not used{ opplicable to CPSSS. S;c conscreion c y ;ri;Gn 3.2 6

                                          -6Bh-
                                                                                        ~

3.2 19 Not applicable to CPSES.7ee w..;;d- prte, duie l (enclosure 68). 3r2:20 Thi.s;changeijecorporatesiTSRillRL(iieMdelstes;the Noteiin o.3.2 1 i Conditioned which? required (that;requirediaction 0.43be Co8Pleted;wheneyeriConditioniDjj s tentered)l.5 Deleting ithe Note? permits:exitingithelrequ1rediactioniwhen;powercis; reduced bel ow;50t@TP TZ Exiting;the!requi red:actionlislappropri ate;because operationibelow 50tJtR41s"not.; constrained;with: respect to:AFD? CPSES Differencesfrom NUREG-1431 - ITS 3.2 4 7/29/98

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g,_--_--__-.___.______.__---__.-_.__---___---_.--________ _ . - - - _ _ _ - ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.2-10 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.2.4 Quadrant Power Tilt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (All FLOG Plants) JFD 3.2-15 ITS SR 3.2.4.2 Comment: JFD 3.2-15 justifies numerous changes to the STS one of which is unacceptable. JFD 3.2-15 is based upon TSTF-109, which has been rejected. The unacceptable STS change is: The modification of the note to SR 3.2.4.2, and in particular the addition of the 12 hour allowance in the Note to SR 3.2.4.2. Provide adequate justification for this change or adopt the STS version of the Note. FLOG Response: The latest status report from the TSTF industry database, dated June 16, 1998, indicates that the NRC has approved TSTF-109. The FLOG continues to pursue the changes approved in TSTF-109. JFD 3.2-15 is revised to delete the sentence: "The note for SR 3.2.4.2 is changed to require performance if one 'or more' OPTR inputs are inoperable." and added: "The note and Frequency for SR 3.2.4.2 are revised consistent with typical presentation formats that provide for a period of time after establishing conditions." NUREG-1431, Rev.1 currently has "or more" in the Note and TSTF-109 did not modify this wording. ATTACHED PAGES: Enc!SA 3.2-16 Encl 6A 4 j Encl 6B 3 1 J l l l l l

OPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 -- --- - - NOTES - -- --- --

1. With input from one Power Range j Neutron Flux channel inoperable and I THERMAL POWER $ 75% RTP, the remaining 13-2-09A i

three power range channels can be used { for calculating QPTR. I l

2. SR 3.2.4.2 may be performed in lieu of this Surveillance if adequate Powcr <3.2-15 Range Ncutron riux channel inputs arc ,

not OPEPJ-SLE. Verify OPTR is within limit by 7 days calculation. M Oncc with4ft '3.2-10 ' 12 hours and cvery 12 hours thereafter with the OPT", alarm inoperable SR 3.2.4.2 - --- - -

                                                                                                                           ----NOTE                             -        - -    -                    03210 Orb Not: required to be performed 4f until;12 hours afterjinput from onQr moreierAp                                                               n3.2-15 ,

Power Range Neutron Flux channeis aie inoperable with THERHAL POWER 9 75% RTP. 3.2-09: Verify QPTR is within limit using the Oncc within 53.2-15< movable incore detectors. 12 hours M 12 hours thereafter 1 i CPSESMark-up of NUREG-1431-ITS3.2 3.2-16 7/29/98 L -_

CHANGE NUMBER JUSTIFICATI.QH 3.2 15 This change incorporates industry traveler TSTF 109. Action A.2 would require the OPTR be determined rather than 0 3.2 10 performing a specific surveillance becna more than one r' surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1Minote"andMency9orJU.2.,.^..cr sed to'be consistent;With; typical [ presentation; format.s that

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                                                 =m" OP3 inputs arc inc erab]p.- TFiese changes are acceptable because tt ey clarify the ISTS regarding frequency and use of incore flL< monitoring for OPTR measurement. The changes reflect that incore detectors provide an acceptable OPTR determination during all plant conditions.

3.2-16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that Fa(Z) is within its limit. Fa(Z) is approximated by Fj(Z) (which is obtained via SR 3.2.1.1) and Fs(Z) (which is obtained via SR 3.2.1.2). Thus, both Fj(Z) and Fg(Z) must be established to verify Fa(Z). This change is consistent with traveler WOG-105. 3.2 17 Not used. applicabic to CPSES. Scc ccavr.rsion comparisca 0 3.2 3 tab 1c (caclosurc 50). 3.2-18 Not applicable to CPSES. See conversion comparison table (enclosure 6B). 3.2 19 Not applicable to CPSES. See conversion comparison table , (enclosure 6B). j i 3.2!20 Thi s change;1. incorporates JSTF;112 L (i . e .",idel etes Lthe; Noteii n 0 3.2 1 Condition;D whic.h; required;that! required; action.D)11be compl eted, wheneverJCondition;Dli s1 entered) C Deleti ng _.the Note; permits: exiting;thelr.equiredfactio[ when power;1s1 reduc.ed below 50% RTP;L Exiting,the; required; action;is1 appropriate because operation;belowl50t1RTP'is;not: constrained;withfespect to AFD; s 1 l 1 i e CPSES Differencesfrom NUREG-1431 - ITS 3.2 4 7/29/98 L____-_-_____________

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                                                         ; ;~    f rr od pl C fFh i oiin sal iT cv e

A r1 rei a - b a _ t d r c , S df ut e o qc mk eeeTh n r cSi e P y M pFd T eep _ eS ncr r _ h e t m o r tt  : c t a m e 1 . eAA r d eth p;we eit ss .h O gTi no - n mad - b o f c n u ~d g nvT r

                                                                       .S n rl ci C             f 1eD aoE and St ~ w C         a rrl                    c   s r       q . i           a e             d f h eel n, cl t e

r p e c r c md hR t c iuel eeP S oit nqv ov sC pt e roe itZ^i N eee s vd vd u n c uh s c s1. rr" ec owl u r o ril sod p O i a h r eus r e q e eh 5 0 c. i i3 h4 uoph eh h noo p es eicnm I S T tb s u R w ' cd

                                                         ;       T1 ct aT rt TiCc R

E V N O R _ C E 5 6 7 8 9 0 B 1 1 1 1 1 2 M - - 2 2 2 2 2 - U Z. - N 3 3 3 3 3 3

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP-3.2-ED APPLICABILITY: CP l REQUEST: Minor corrections / consistency changes l l l ATTACHED PAGES: Enci5B B 3.2-14 Consistency change to replace " steady state" with " equilibrium" B 3.2-29 Corrects a FSAR reference

l Nan B 3.2.2 BASES i provide the oeactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, a;. 'NB events in which the calculation of the 0 3.2.G 1 core limits is modeled implicitly use this variable value of Fin in the analyses. Likewise, all transients that may be DNB limitedareassumedtobeginwithaninitialFin as a function of power level defined by the COLR limit equation. The LOCA safety analysis ir,directly acdcis als_oluses Fin as an input parameter. The Nuclear Heac Flux Hot Channel Factor (Fo(Z)) and the axial peaking factors are inserted directly into the LOCA safety analyses that verify the acceptability of the resulting peak cladding temperature (RefC3). 0 3.2.G-1 The fuel is protected in part by compliance with; Technical Specifications which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LC0 3.2.3, " AXIAL FLUX DIFFERENCE (AFD)," LC0 3.2.4, " QUADRANT POWER TILT RATIO (QPTR)," LC0 3.1.7,

                       " Control Bank Insertion Limits," LCO 3.2.2, " Nuclear Enthalpy Rise Hot Channel Factor (Fin)," and LC0 3.2.1, " Heat Flux Hot Channel Factor (Fo(Z))."

Fin and Fa(Z) are measured periodically using the movable incore detector system. Measurements are aenerally taken withthecoreat,ornear,pady;tateequil.ibriup CP 3.2.ED conditions. Core monitoring and control unaer translent conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits. Flu satisfies Criterion 2 of the N!10 "clicy Statc;cr.t 10CFR50.36(c)(2)(ii). LC0 Fin shall be maintained within the limits of the relationship provided in the COLR. The Fin limit is . representative of idertifics the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB condition. l (continued) CPSESMark-up of NUREG-1431 Bases-IIS3.2 B 3.2-14 7/29/98 l

                                                                                                                    \

i i AFD (CAOC Methodology) B 3.2.3 i BASES (continued) l REFERENCES 1. WCA" S403 (nonproprietary) "Pcwcr Distributica Control and l Load following Procedurcs Westinghouse Electric I l Corporation. Septccber 1074. RXE.90 006 P A;" Power l Distribution: Control) Analysis and:0vertemperature;N 16:and Overpower.N ~16l Trip.Setpoint Methodology,"iTV Electric;; June,E199.4 l

2. WCAP-8385;(W proprietary);.," Power Distributionm Controlland l Load Following Procedures," Westinghouse Electric Corporation;LSeptember 1974.'

1 3 ;' T. M. Anderson to K. Kniel (Chief of Core Performance { Branch, NRC),

Attachment:

" Operation and Safety Analysis Aspects of an Improved Load Follow Package,"                         ]

l January 31, 1980.

3. C. Eicheldingcr to D. B. "assallo (Chief of Light "ater -

Reactors Branch. N",C), Letter NS CE 507, July 15, 1075. l l 4 FSAR, Chapter {-15 7 lCP3.2hD I l l I CPSES Mark-up of .YUREG-1131 Bases -ITS 3.2 B 3.2-29 7/19/98

l 1 ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: CP-3.2-001 APPLICABILITY: CA, CP, DC, WC REQUEST: The mark-up CTS SR 4.2.5.3 (4.2.3.4 for DCPP) was revised as follows. This SR requires a CHANNEL CALIBRATION on the RCS toop flow rate once per 18 months (refueling interval for DCPP). The CTS SR is equivalent to ITS SR 3.3.1.10 and the Reactor Coolant Flow - Low functional unit. DOC 5-12-A was initiated to address that CTS SR 4.2.5.3 (4.2.3.4 for DCPP)is equivalent to ITS SR 3.3.1.10. For CPSES and DCPP, the strikeout is removed consistent with the FLOG markup methodology. For CPSES, the CTS SR 4.2.5.3 statement "The channels shall be normalized based on the i RCS flow rate determination of Surveillance Requirement."is struck through and DOC 5-03-LG I applied. DOC 5-03-LG is revised in Enclosures 3A and 3B to indicate the DOC is applicable to ) CPSES only and that this information is moved to ITS Bases 3.4.1. ' ATTACHED PAGES: EncI 2 3/4.2-12 Enci 3A 8 & 10 Encl 3B 6&8

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the stated limits:

a. Indic ted Reactor Coolant System T,ys 592"F l05a01 lgs

b. Indicated Pressurizer Pressure 2 2219 psig* (05.01;LGA

c. Indicated Reactor Coolant System (RCS) Flow g 403.400 gpm** for Unit 1 E' 05 01 LG3 2 408,000 gpm** for Unit 2 APPLICABILITY: MODE 1. I ACTION:

I With any of the above parameters exceeding its limit, restore the parameter to I within its limit within 2 hours or reduce THERMAL POWER to less than 5% of I RATED THERMAL POWER within the next 4 J hours. . {

                                                                                                                                     !05.06 LS:        J With RCS' flow ~ measurement:per;Specif1 cation 4.2;5;4 not meeting 11a.it,ido not       .

exceedj85(RTP;

                                                                                                  ~    ~ '               - - -
                                                                                                                                       ;05 11 A.a SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the above parameters (RCSlT, Z.. Pressurizer Pressurei                !!05-11-A ?

indicated;RCS flow) shall be verified to be wlthin its limits at least once per 12 hours. 4.2.5.2 The RCS total f1;w retc -hall be verificd t; bc within it:; li;;;its at 105 02 LS3 least enn pcr 31 t ys by plant ca.pter indication er measurc; cat of the RCS

                                                                                                                                       ~

alb;w top diffcrcatial pressurc tren,;;;itters' cutput volt;;e. r 4;2iS137The RCSiloopiflow;rateMndicatorsishallibe(subjected _to.a; CHANNEL 05 12 A CALIBRATIONlat l east oncei per '18j months !:4. 2. 5. 3 The RCS leap flow rate CP-3.2 001 indicater hell ',a :ub.iected t; ; C"^fM' C'LI".R' TIS et l=:;t cree gr la

                                              - -n " - 'The chennels sh ll be nor;;;;lind b; sed on the RCS flew rh                                  -

ger;;;ination of Survcillerec Rcquir;2nt 4.2.5.4. 05 03 LG CP.3.2-001  ! 4.2.5.4 The RCS total flow rate shall be dctcr;;;incd by precision heat i bel;nce ;xsurc;ent measured after each fuel loading and prior to operation above 85% of RATED THERMAL POWER. The fuieter prasur =d to..prature. ,1 05 MG o, . i the ;;;ain stc= pres:;ure. and fGieter flow differcatial pics:;ure instrunnts shall be celibrated within 90 ty:; cf perfor;;;ing the calori;;;ctric fisw masur;nnt.

• Limit not applicable during either a THERMAL POWER ramp in excess of 5%

of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER. Includes a 1. t f1;w == sur;x nt una rtainty. r05 01-LG? CPSESMark-up ofCTS 3M.2 3M 2-12 7/29/98

CHANGE NUMBER HSHC DESCRIPTION 04 07 A Not applicable. See conversion comparison table (enclosure 38). l 04-08 Not used. 04 09 A Not applicable. See conversion comparison table (enclosure 1 3B).  ! i 04 10 LS-14 Not used. applicabic. Scc ccaversica comparison table I (enclosure 3Sh. 0-3.2 6 04ill A Not applicable; JSee: conversion comparison table _ (enclosure 38), _ 05 01 LG The designation of how instrument uncertainties are treated (nominal, in the analysis, or in the development of the TS limit) is moved to the Bases. The movement of this level of detail out of the specification is corsistent with NUREG 1431 and is an example of removing unnecessary details from the TS in accordance with 10 CFR 50.36. 05 02 LS 7 The requirement to verify that the total RCS flow is within limits using the plant computer or elbow tap output voltage on a monthly basis is deleted. This action was included in the TS based on the potential drift of the RCS flow indication. The essence of this activity is performed on a quarterly basis through the Channel Operability Test performed in accordance with the Reactor Trip System surveillance. _ . 05-03 LG Ccasi m .,, ,,,,,,,m.u.,,~. , , ,m

                                                                              . w o . . m.om . ' ' 'v                         CP 3.2 001 peric,a a C"ANNEL CALIBRATION cn thc RCS flow meters et least once per 10 acnths and The requirement to normalize the RCS _1_oop flow rate indiators J. channels are is to the Bases for the RCS ficw low reactor trip fu. m Q TS Section 3.34.1.

05-04 LG Consistent with industry traveler TSTF 105, the explicit requirements that the RCS flow be measured through the use of a precision heat balance measurement and that the instrumentation used in the performance of the calorimetric j flow measurement be calibrated within a specified time period of performing the measurement is moved to a licensee controlled document. The requirement to verify that the RCS flow is within limits remains within the Technical Specification. This is an example of removing unnecessary details from the TS and is acceptable based on the guidance provided in 10 CFR 50.36. l l CPSES Description of Changes to CTS 3N.2 8 July 29,1998

CHANGE NUMBER NSliC DESCRIPTION under administrative procedures and will be included in the Bases of the Improved Technical Specifications are:

1) ineasured RCS flow based on elbow tap differential pressure measurement prior to Mode 1 is within 20% of i the expected RCS flow:

l

                                                                                                                                              )

2) the power dependent enthalpy rise peaking factor (F,)

has been verified to be within its limits; and

                                                                                                                                              )

3) the trip setpont of the power range neutron flux -

high reactor trip function is maintained at a reduced setpoint (90% RTP) until the RCS flow has been verified to be within analyzed values. In addition, the parameters to 5e verified per the LCO are clarified. Failure of the precision flow measurement when below 85% RTP following a refueling outage does not result in the violation of the LCO: it only prohibits power ascension above 85% RTP. Sil2 A Thelpequi.rement to; perform [18] month CHANNEL i CP 3.2 001 CALIBRATIONS.lof1the RCSfloop'flowrateiindicatorsijs part offITS sri 3;3;1;10;for Reactor 1 Trip l System Inst rumentati.on: Function : 10 : (Reactor 1 Cool ant l Flow!$ Low) . l CPSES Description of Changes to CTS 3N.2 10 July 29,1998

A f L o L s s s s N A s s 8 C Y e Y e Y e A N Y e N o e Y N o V e f o 6 - S g T 0 C e g 1 n a K 4

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                                                          -                                                         N O          -

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                .od          n             r                            et s             rh            ao               ch oi 0           sPl r          u i

m ye l p oa t wcr ou u t t tb ol i t t f S ut o i po tl Pd o ef t e t edS I oew l e p u . er eh r o a er a z eC v R h s e% e ad e d c% el 5 h t r7 2 s n t h o O" i l a mc o" A 4 r5 2 o 5 b t u on d a7 d c a e ee t r I m h P d ni t o 5 t wo eot i m t p t e . atd T rst M int < l sn st s s p no h ue L" . 1 n oi r hd o u eo e d t pt " co O t nt o oc i eer d i rge il ae me I orf . C i a w ,tn w d t t s e oin nd t v r e yol f ce ' L t a c1 9

4. m i ouw t r e ed i d C t. c :

d ri 0 N o ut c rt .i n4 N eeo r wp 0 r e ep td x ne ens l ~ . ce3 O E 1 ab n e uh vai n i G oot 3 p d et mne t l I .r a- s n cs I N t pe s s g n ns 1 on

                                                             . t a       eib t

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                                                          -         n h i N tHO I             wye s         u un q         o f

rddl t i ai C S T g P n - T N P pi S vd G E r ei r - T ol h h at esm rt c nnl c paii n H Oi Oh eo R os o C n i a c r x t t r i U ci f xA f a o cm o;evf C guu o ai a r N x u t i i n t h t r E ndl n w h s e el e el R e f l o t. a up N i f i t ap h mFT nhd i t . rci T rd ve tE n i P ot e cn e O I i uio en l c on tdd n ee iL n wB a t nQ o i t no v eig ph a newet e m oh T qf r . va ed t Ah d r y . ai m st t eCl t r _ P eit d nm mei tR c et bk n i l rI f rrue i r eem nE wu e g .s S wo iu - pocrt I eet o t ni d eP e oed e i l i E v R s vna n sf a l e t O n l N er s a S s qcof a n nn t rn s s o l rC en) Pi t ep o C oeei oo seu U i e a ei dit C wp u r l sr S ib gm i c g s en guf mi c e E t ni t en r t nr e enql eom eot e S s ; D coal cn h oo N o ohh h a eo h ni hl u h . c. C a At r e Ai Tl f Ct w TR rW T( l Tf o T uRBi R E _ B 51 63 7 8 9 04 1 2 3 M _ 01 01 0 0 0 1 1 0 07 0 U 4S 4S 4 4 4 4S 5G 5S 5G N 0L 0L 0A 0A 0L 0 L _ 0 0L 0 L - l ll!

A L t e L .tp A s a s s eh e o e 8 C YC Y N Y r o 8 e t g n K e a m p E R A . e r . E S6 iS R U1 uT Y C or qC e T F t e R n I L ,tp - i L O sa t s I eh oo o e B W YC N n N Y A C t t I L K o o n n P A P E t n t n 2 A P e e m m

       . E   e          e r           r 4        H C

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A RS RS M T C T S O - C s s T C on Ni on Ni Y e Y e T N N E O t R Y n e R N m U A C e. r C iS uT O qC e

     -      L R n B       i E        A I

t s s L oo e o e D N n Y N Y B A r S T r e o i t r sC iR . e r nl t S N n t rao h s aeu O o N a w p d f s t n o t s O we h s I T i t 0 d e sns t aR i di ns i x C Lt r r AR c S ed i mg S nt ai e nt e R5; B nu I f e or c eeow I SF l i ne i o ,e w h mno LT R el h o l i h dt t per o t e rs p AI n. A t r na cpt pl aua sra f rues C 2 o f i P gn t i om hl o n d es ot padi Lot E;a M no i c t il f r A swf e eoor e d mei gh b N tt O n a e we .sl i e N rn A a e C ee s ef e ud kl yf eu r q i wa o f ot r H pm C _u l cn t ah a t a ad ue enel r i l up sr rf o hit N e rcn P aci e y t ;s O E d c ai uae v n gl nsn G ni wem i etdf h onn orI I al o h ne t eiio mo;? S N go l f ee rl cere nd h t b sl .t m A i et ] ae) R H i nt Sf e t aiat a mc h ocui t ef 8 ct w 1i so E C td ce C ab R och yt C atb r e: s pro [d yL nS V C ev po yTl Pd i y r C mi;X r;p N E s m f R u oewdf th ei eeaL th . oeiw O P n i r% w o c tPt r T ce et geP m nhT ft ro raTl _ C S i rb e e5 v7 s i t R ev f a f af itR r owof% er:F pwr H o D ih% f o a . oot f d oeF ct 5 at pel o5 ol t n C l t vP e py 8 rl 8 tf ca E su oE tt euon  ; al t o tb sf won hl f oe t peo T NO nw e na1 s e 3 r s rl i one m t i aP t o i v n o R. o e m1 or ; C, _ I T P mi er ru it s mn eoyt rire i t em e i a ese ec g l bi r nns u n

                                                                  ,FT ab o . Rlo a i d% i n e'; o r rSf o iC;t I   unt qen ua qr er va          noooa              t e5 vo             uR0 c q

R at ii e i i 8 i 1 a C eve ee a h t t m df es ee e r m di wh n r th 1. R( rpt p cd ai S eoc wu e osaB erd w a rot e E ed eno hd pol al c e h f" ;0 3' hl o h oeN nl ens D Tfd Ttl D T aocf I cbi a T o31 R E _ B 9 0 1 2 M 0 1 1 1 U 5G 5 5 5' N 0L 0A 0A 0A

I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR-3.2-004 APPLICABILITY: CA, CP, DC, WC REQUEST: Update traveler page for latest approval status and revision numbers. ATTACHED PAGES: Enci5A traveler page l l i l 1 I i i

                                                                                      )

INDUSTRY TR2WELERS APPLICABLE TO SECTION 3.2 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-24 Not Incorporated NA Not NRC approved as of traveler cut off date. TSTF 95 Incorporated 3.2-06 Approved by NRC TSTF 97 Incorporated 3.2 07 Approved by NRC i TSTF 98 Rev. fl Incorporated 3.2 03 TSTF 99 Incorporated 3.2 08 Approved by NRC TSTF 109 Incorporated 3.2-15 Approved by NRC TSTF 110 Rev. -1 Incorporated 3.2-10 M ved by,NR .m; .2 004 w_ f%- l TSTF 112. Rev. 1 Net Incorporated NA 3.2 20 N^t NRC apprev-d 0 3.21 as of traveler cut-off date. Approved by NRC.? Applicable to CAOC plants. (CPSESionly) Incorporated TSTF 136 NA ov 3 lT .5004 TSTF 164 Incorporated 3.2-11 Applicable to CAOC only. (CPSES) WOC 05, proposed Incorporated 3.2-05 0 3.26 Rev. 2 TSTF 241 3.2 189 Rev;21 W0r r 105 Incorporated 3.2-16 i

i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC-3.2-001 APPLICABILITY: CP,WC l REQUEST: CTS SR 4.2.4.1 and 4.2.4.2 markups for Wolf Creek did not include ITS SR . i 3.2.4.1 Note 2 and SR 3.2.4.2 Note. Adding these Notes to the CTS markup resulted in ' creating DOC 04-11-A which is only applicable to Wolf Creek is maintaining the CTS requirement for performing SR 4.2.4.2 with only one Power Range channel inoperable and not adopting the changes proposed in DOC 04-04 LS-12. For CPSES, CTS SR 4.2.4.2 was not marked up to include iTS SR 3.2.4.1 Note 2. The Note has been added. The addition of the Note is covered by existing DOC 04-04-LS-12. ATTACHED PAGES: Enci 2 3/4.2-11

f, - POWER DISTRIBUTION LIMITS l SURVEILLANCE REQUIREMENTS l 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: _ 04 04.Ls WC 3.2-001

a. Calculating the ratio at least once per 7 day 6, when the clara is OPEPJ"LE

b. Calculating the ratic at least cacc pee 12 hours when the alara !01 07.-LG >

is incparabic. and

c. Calculating the ratio at least once per 12 hours

• when above 75%

RATED THERMAL POWER with one or_more Power Range' Channel (s) 104 04 LS-inoperable.

• Notirequired'u_nt11'12 hours l.after.,(input;from.one orimore Power Range 5: 04-04 LSa Channel (s);become;jnoperable'.

        **; Surveillance. Requirement 4.2:4'.1c may..be performed;in:11eu;ofuthis surveillance.: requirement?                                                                                                                  0404.ts\

WC-3.2 001 1 CPSESMark-up ofCTS 3N.2 3N 2-11 7/29/98

Attachment 3 to TXX-98182 Page 1 of 5 l JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS CTS 3/4.5 - EMERGENCY CORE COOLING SYSTEMS ) i ITS 3.5 - EMERGENCY CORE COOLING SYSTEMS l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES l

Attachment 3 to TXX-98182 Page 2 of 5 INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED j NUMBER 3.5.G-1 DC, CP, WC, CA YES 3.5.1-1 CP YES 3.5.1-2 WC NA 3.5.1-3 CP YES 3.5.1 -4 DC NA 3.5.1-5 DC NA 3.5.1-6 DC, CP, WC, CA YES 3.5.2-1 DC, CP, WC, CA YES 3.5.2-2 DC, CP, WC, CA YES 3.5.2-3 DC, CP, WC, CA YES 3.5.2-4 DC, CP, WC, CA YES 3.5.2-5 DC, CP, WC, CA YES 3.5.2-6 DC, CP, WC, CA YES 3.5.2-7 CP YES 3.5.2-8 WC, CA NA 3.5.2-9 DC NA 3.5.3-1 DC, CP, WC, CA YES 3.5.3-2 DC, CP, WC, CA YES 3.5.3-3 DC, CP, WC, CA YES 3.5.3-4 DC, CP, WC, CA YES 3.5.3-5 DC, CP, WC, CA YES 3.5.4-1 DC NA 3.5.5-1 DC,CP YES 3.5.5-2 WC, CA NA CA 3.5-001 DC, WC, CA NA CA 3.5-002 DC, CP, WC, CA YES CA 3.5-003 CA NA l CP 3.5-002 CP YES

l. ' CP 3.5-003 CP YES .

CP 3.5-004 CP YES _____ ____a

Attachment 3 to TXX-98182 Page 3 of 5 INDEX OF ADDITIONAL INFORMATION (cont.) ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER DC ALL-002 (3.5 chages only) DC NA DC 3.5-ED DC NA DC 3.5-001 DC, WC, CA NA DC 3.5-002 DC NA DC 3.5-003 DC NA DC 3.5-005 DC NA DC 3.5-006 DC NA TR 3.5-001 DC, CP, WC, CA YES WC 3.5-ED WC NA WC 3.5-001 WC NA WC 3.5-002 WC NA WC 3.5-003 WC NA l l l i

l l l Attachment 3 to TXX-98182 Page 4 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:

1. Each licensee is submitting a separate response for each section.

2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

3. If a licensee initiated change does not apply, 'NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,38, 4,6A and 6B of the conversion submitta's). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as ' PLANT SPECIFIC RESPONSE ... . ."

5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.

6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.

7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one or more of the other licensees. These I changes are not included in the additional information for the licensee to which the change does not apply, as the changes are only for consistency, do not technically affect the request for that licensee, and are being provided in the additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment l request supplernent to be provided later.

{ l t l i L_______-__--_-_______ . _ _ _ _ - _ . ]

I Attachment 3 to TXX-98182 Page 5 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued) I

8. The item numbers are formatted as follows: [ Source][lTS Section]-[nnn]

Source = Q - NRC Question CA- AmerenUE DC-PG&E WC-WCNOC CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL"is used for the section number. nnn = a three digit sequential number or ED (ED indicates editorial correction with no impact on meaning)

ADDITIONAL INFORMATlui. COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.G-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.5.x Bases General There have been a number of instances that the specific changes to the STS Bases are not properly identified with redline or strikeout marks. Comment: Perform an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal. FLOG response: The submitted ITS Bases markups for Section 3.5 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorialin nature and would not have affected the review. Examples of editorial changes are:

1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced.

2) Changing a verb from singular to plural by adding an "s" without " redlining" the "s".

3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g.,

SR3.6.6A.7).

4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference.

5) In some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.

6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.

7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was

                     . determined to not be applicable, the text was then struck-out and remains in the ITS Bases mark-up.

Differences of the above editorial nature will not be provided as attachments to this response. The pages requiring changes that are more than editorial and are not consistent with the markup methodology are attached. ATTACHED PAGES: l EnclSB B 3.5-31 one bracketed number not relined and another bracketed number not struck-out l E_--_---_____ .1

Seal Injection Flow B 3.5.5 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.5 Seal Injection Flow BASES l BACKGROUND This LCO is applicabic caly to the:c units that utilizc;the centrifugal charging pumps for safety injectica (SI). The function of the seal injection throttle valves during an accident is similar to the function of the ECCS throttle valves in that each restricts flow from the centrifugal charging pump header to the Reactor Coolant System (RCS). The restriction on reactor coolant pump (RCP) seal injection flow limits the amount of ECCS flow that would be diverted from the injection path following an accident. This limit is based on safety analysis assumptions that are required because RCP seal injection flow is not isolated during SI. APPLICABLE All ECCS subsystems are taken credit for in the large SAFETY ANALYSES break loss of coolant accident (LOCA) at full power (Ref.1). The LOCA analysis establishes the minimum flow for the ECCS pumps. The centrifugal charging pumps are also credited in the small break LOCA analysis. This analysis establishes the flow and discharge head at the design point for the centrifugal charging pumps. The steam generator tube rupture and main steam line break event analyses also credit the centrifugal charging pumps., but are not li;;;iting in their dcsign. Reference to these analyses is made in assessing changes to the Seal Injection System for evaluation of their effects in relation to the acceptance limits in these analyses. The ECCS) flow balance; assumes;RCP' seal;injectionds:limitedLto30 ) gpm with.FCVil21Lfu11 openf andl centrifugal; charging l pump headertat 130(psig1origreate Ethan the Reactor;CoolantLS ystem pressurei(1;er; the; pressurizer);  ! This LC0 ensures seal injection flow of shpm with l cer""ugal charging pump discharge header RCS pressure Q 3.5.G-1 2400 215 psig and < 2255 psig and charging flow control va full open, will be sufficient for RCP seal integrity > but limited so that the ECCS trains will be capable of delivering sufficient water to match boiloff rates soon enough to minimize uncovering of the core following a large LOCA. It also ensures that the centrifugal charging pumps will deliver sufficient water CPSES Mark-up ofNUREG-1431 Bases - ITS 3.5 B 3.5-31 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.1-1 APPLICABILITY: CP REQUEST: DOC 1-02 A JFD PS CTS 4.5.1.8 STS SR 3.5.1.4 ITS SR 3.5.1.4 The STS wording for the second Frequency of SR 3.5.1.4 provides an option for specifying number of gallons or percent of indicated level. ITS SR 3.5.1.4 uses percent, but the phrase "of indicated level" has not been adopted. This change is not consistent with Diablo Canyon, the other licensee who chose to use percent. Comment: Make iTS SR 3.5.1.4 consistent with the STS and with Diablo Canyon ITS SR 3.5.1.4 by adding the phrase "of indicated level" or provide a plant-specific reason for not adopting the STS wording. FLOG Response: ITS SR 3.5.1.4 will be revised to be consistent with the STS. In the original submittal (5/19/97) CPSES chose to use a "%" level instead of level specified in " gallons" as was contained in the CTS. We have now reconsidered and have chosen to specify level in " gallons." DOC 1-01-M and DOC 1-02-A are no longer required for this change. Also see response to 03.5.1 3. ATTACHED PAGES: Enci 2 3/4 5-1 and 2 Enci 3A 1 Enci 3B 1 Enci5A 3.5-2 Enci5B B3.5-4 and 8

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITED CONDITION FOR OPERATION 3.5.1 Each cold leg injection accumulator shall be OPERABLE with:

a. The discharge isolation valve open with power removed. -

4-M-H l

b. 0 3.5.1 1 A contained bor Medger volume of between 511 ,E611 'and 65 614 6597 gallons allcr.3, c.

YW A boron concentration of between 2300 ppm and 2600 ppm, and

                                                                                                                                                                                       \
                                                                                                                                                                                       )

l

d. An nitrogen cover pressure of between 603 623 and 693 644 psig, s i.et.ne a> )

APPLICABILITY: MODES 1, 2, and 3*. i ACTION:

a. With threelaccumulatorsj0EERABLEEand one cold leg injection accumulator t i.es.A q inoperable, except as a result of the boron concentration outside the required values, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and [ M As reduce pressurizer RCS pressure to less than 1000 psig within the following 6 hours.

b. With threetaccumulatorsiOPERABLEland the boron concentration of one ,,g cold leg injection accumulator outside the required limit, restore the boron concentration to within the required limits within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce picssurizer  ! 14A RCS pressure to less than 1000 psig within the following 6 hours.

                                            *l'rassurizar RCS pressure above 1000 psig.                                                                                   il4AN l

CPSESMark-up ofCTS3N.S 3 4 S-1 7/29/98 1

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS l 4.5.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: )

a. At least once per 12 hours by:

1) Verifying that the contained borated water volume and nitrogen cover-pressure in the tanks are within their limits, and

2) Verifying that each cold leg injection accumulator isolation valve is open, i

b. At least once per 31 days and within 6 hours after each indicated -

solutinn unlume increase of areater than or equal to 101 galicas m

                 @l: gallon)y verifying the boron concentration of the                      0 3.5.1-3 soiution in tne water filled accumulator. This surveillance is not required when the volume makeup source is the RWST and the                r wi-w . 1 RWST has not bcca diluted sinc; verifying that the RZ T bcron concentration i:; equal to or grcater than the accumulated boren coaccatration limit.

c. At least once per 31 days when the RCS pressure is above 1000 psig by verifying that power to the iso 1*ation valve operator is removed.

l l 1 CPSES Mark-up ofCTS 3N.S 3 M S-2 7/29/98

DESCRIPTION OF CHANGES TO CURRENT TS SECTION 3/4.5 This enclosure contains a brief description / justification for each marked-up change to the current Technical Specifications. The changes are identified by change numbers contained in enclosure 2 (Mark up of the current Technical Specifications). In addition, the referenced No Significant Hazards Considerations (NSHCs) are contained in enclosure 4. Only technical changes are discussed: administrative changes (i.e., format, presentation, and editorial changes) made to conform to NUREG-1431 Revision 1 are not discussed. For enclosures 3A, 3B, 4, 6A and 6B, text in brackets "[ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUMBER [ gig DESCRIPTION 1 01 H In accordancc with NUREC-1431 Rcv 1, the : indicated tank icycl is added to clarify the volumc. In .

                                "dition/ffieTantsoecificparametersfo accumulator 6cnk icvci ard) tank pressure ere:is                0 3.5.1 3 revised to provide hmits that can be directly confirmed using control room instruments. ThisLis a more. restrictive; change becausedit reduces the: allowable pressure' range for the accumulatorjcover pressure. There are;actually two changes-being made. "The first change corrects _'an existing error in the_ units in the CTS.L The existing CTS; pressures,(603;and'693)1are; psia: numbers not psig as; indicated in the CTS. 3The correct psig numbers z

would be 588:and 678.;This results .in the _ higher. CTS 4 pressure;being non conservative.~The existing implementation; procedure;for, the accumulator; pressure had the: correct pressure..r.ange; The second change' applies instrLeent error:to the' analysis; pressures 1resulting in newl pressure range..623;psig to:644 psig. lThis change j allowstheuse:of: control 1roomjevel;indicationtoperform i the surveillance; Thcsc. changes arc. administrative as therc are no technical differences in thesc numbers. Therc arc as technical changes to the analyscs. 1-02 A Consistent with NUREC 1431 Rev 1. the nominal tank 0 3.5.1 1 i vcluac incrcasc of 101 galicas is clarified to bc lt i of tank icvci which is the parameter acnitored by the operator. This changc is administrative as thcrc arc no l technical differcaccs in the two numbers. Not used.' i l l i ' 1 03 A Replaces reference to the " pressurizer pressure" with a j reference to the "RCS pressure". ACTIONS a and b require ' reducing pressurizer pressure to less than 1000 psig. However, pressurizer pressure instrumentation does not l have the range to read that pressure. Consequently RCS l pressure instrumentation is used. For the purposes of I this LCO, the use of RCS pressure is equivalent. I CPSES Description of Changes to CTS 3N.S 1 7/1988 L ]

P A C ni dS A e L L

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   -   C Y                 V             Y        N e

N Y Y E L B N d O e A Y N i n i T A t a e d y C N O S r a c r O L T C " A N S B I A - S s s s s s R I D N o o NG Y e e Y Y e Y e Y e A P M c . oo

           - t r l

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un g E L 0 n e T ab t o er m d - co cit f f R l c - i cc c o s e r ct aat o cc oo a t B A R 3 . l l s ee i a rnt pm". E o r n ef i r v p l f oeo n n f o n t o n E O e P Cl eh v nt e V t uo s c ci r c r os mi or ni O Lb a rT oa e i et o et uS i t N E d s r z n r c nb mc s or sW t s g G c ey d rl ica i r odi u i eud o i n eu rd rt o ne p r R a r ei n O N d ct a c c m u t s qe aoer t a i e ur t i on ue th y C A H k e c n r c c c s". s e e rl r oc e ah rt o q ee l a u r yi oh nt f h t e cer i C i ai ch r r t er t r rr mt s 6d nv e l td -

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                             - c         t p f

ot a o os d f n no 3 amm lu rW rui iR oos k cc ui oS rst oat h u u b w H n y l t t C eer es t oc qe d a ct o a mh ud ma rd it c eh see C t i a vc e R i t oe i eue w a rt t gt E h . i c' ne t oh g n t rog " s ys i n u T N dk t s c n t k d n n eh nr2 a nr h n a t e s n r l i ghi al O t a s n a' rt oo7h ooI h d eo t ncd . I T ct t e ic r imum t l h e f o et if t cyt sg o c . icytfs go t c. rmm oe wt r nei ecyd ne r rf ees s P d oir at r At i i At i i a o r uis e a I ntl t n e id s id s et uor obB R " a s if sc Pl eN p Pl eN p h so cs ep C l en - o en Pi sW P i sW t w v ot4 t udi o na ce a r Cb aO0 Cb aO0 nt R p ro S mi uv m e D aeD0 D a eD0 d o S uepn1 E t l rrT 0 rrT0 eih ec D ecoo ci pf eecU1 eecU1 dt t ek nes5 h c r o T apr h n ee h pnH h pnH d ci h aih a T u R r T oiS 5 T oiS 5 AA w T msTh3 R E B M 1 0 2 0 3 0 48 0 - 59 6 7 U - - - - S 0 -

                                                                        - S 0

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                                                                                                       - G N  1 M              1 A           1 A      1L                  1 L               1 A          1 L

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is 12 hours fully open. SR 3.5.1.2 Verify borated water volumn in each accumulator 12 hours  : B'PS " is 2 7853 g;Mensd9E6119Tgallons3nd

                        ~

0 3.5.1 1 s;8171 gallon Q 6579; gal _ SR 3.5.1.3 Verify nitrogen cover pressure in each 12 hours :B-PSL/ accumulator is 2 385 623 psig and s 481 644 psig. SR 3.5.1.4 Verify boron concentration in each accumulator 31 days !B-PS-is a 1900 2300 ppm and s 2100;260.0 ppm. AND

                                                                      ....-NOTE -          -

Only required to be performed for affected accumulators Once within 6 hours after l each solution volume increas of a gallons J ?B-PS cf in"-"ed 101 gallon ithat ~ 0 3.5.1 1 L. the result of addition from the refueling water storage tank CPSES Mark-up ofNUREG-1431 -1TS 3.5 3.5-2 7/1988 1

Accumulators B 3.5.1 BASES APPLICABLE water volume is the same as the deliverable volume for the SAFETY ANALYSES accumulators, since the accumulators are emptied, once (continued) discharged. For small breaks. an increase in water volume is a peak clad temperature penalty. For largc brcoks pepending;on;the NRC-approved: methodology _usedito;analyzejarge. breaks,an increase in water volume een may;be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequert spill through the break during the core reflooding portion of the transient. The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volume from the accumulator to the check valve. The safety analysis assumes values of E6468} 6119 gallons and f6879} 6597 0 3.5.1 1 inaccuracygallop;Tc nd_e At ;% a;;cw for e :,:; Jeler;nce. ins:

                                                                  ;entEclire;; .

indi;;,t:y f valucs of [5520] geHons and [5020] gallons 30Gr.d 51* are 3-^cified and 3;;dfjn autveill;r;;, j The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent rec'uction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. The large and small break LOCA analyses are performed at the minimum nitrogen cover pressure 1(603.. psia), since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit (693 psia)2 prevents accumulator ! relief valve actuation, and ultimately preserves accumulator l integrity. To; allow foriinstrument inaccuracy,fcontrol_ room 11.ndicated values l of;623;psigfand 644:psigtareLspecifiedsand;used in(surveillance l (continued) CPSES Mark-up ofKUREG-1431 Bases -ITS 3.5 B 3.5-4 729/98

1 Accumulators B 3.5.1 BASES SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day Frequency is adequate to , identify changes that coula occur from mechanisms such as ' stratification or inleakage. Samplin _ accumulator within 6 hours after l'gallonTJolume a 1 @g the affected o.3,5,1 1 increase caused a[reaucuun iiiboron concentration to Delow tne01Tgal required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), anditheLRWST;has not been diluted:since; verifying.that its; boron 1 concentration; satisfies SR;3;5 4;3Ebecause the water contained in the RWST is nominally within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG 1366 (Ref. 5). SP. 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator when the pressurized RCS pressure is > B000 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single i failure coincident with a LOCA. Since power is removed under j administrative control, the 31 day Frequency will provide adequate assurance that power is removed. This SR allows power to be supplied to the motor operated isolation valves when pressurizer RCS pressure is 5 0000 1000 psig.--thtts cliswing operational ficxibility by avoiding unaccessary delays te manipulate the brc;kcrs during plant startups or shutdowns. Even with powcr supplicd to the valves. inadvertent closurc is prevented by the RCS pressurc intericck :ssociated with the valves. Should closure of a valve occur in spite of the interlock, the SI l signal provided to the valves would open a closed valve in the l event of a LOCA.

                                                                                                               )

(continued) l CPSES Mark-up ofNUREG-1431 Bases -ITS 3.S B 3.5-8 7/29/98 l t _ - _ _ __ __m

E ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.1-3 APPLICABILITY: CP REQUEST: DOC 1-01 M CTS 3.5.1.b & 3.5.1.d DOC 1-01M states, "These changes are administrative as there are no technical differences in these numbers." If the changes are administrative, then why are they classified as more restrictive? Also, it is not true that there are no " technical differences" in the numbers. The pressure range changes from 603-693 psig in CTS 3.5.1.d to 623-644 psig in ITS SR 3.5.1.3. Comment: It appears that the two changes captured by this DOC should be separated. The change to the requirements for borated water volume should be an administrative change since it only involves a change in units of measurement. The change to the nitrogen cover pressure requirements appears correctly classified as a more restrictive change, since the range has gotten tighter, assumina an explanation of why the tighter range is necessary can be provided. Othetwise, the changs 7uld be considered out of the scope of the conversion. FLOG Response: The portion of DOC 1-01 M which addressed the change in accumulator level units units has been deleted (See response to 03.5.1-1). The revised DOC 1-01-M addresses the tighter range app]ed to the nitrogen cover pressure. There are two changes being made here which we agree can be classified as out of scope, but we believe are necessary. The first change corrects an existing editorial error in the units in the CTS. The existing pressures 603 and 693 are psia numbers not psig as indicated in the CTS. The correct psig numbers would be 588 and 678. This results in the higher CTS pressure being non-conservative. The existing implementation procedure for the accumulator pressure had the correct pressure range. This editorial error was processed under the CPSES corrective action program and was discussed with the NRC Project Manager for CPSES (Mr. T. Polich) on August 30,1996. The second change applies an instrument accuracy correction to the ana;/ sis pressures (resulting in new pressure range 623 psig to 644 psig) to allow the use of control room instruments to perform the surveillance. This change was missed in our review to identify out of scope changes in the ITS conversion submittal. ATTACHED PAGES: Enci 3A 1 i f t

DESCRIPTION OF CHANGES TO CURRENT TS SECTION 3/4.5 1 1 This enclosure contains a brief description / justification for each marked-up change  ; to the current Technical Specifications. The changes are identified by change l numbers contained in enclosure 2 (Mark up of the current Technical Specifications). I In addition, the referenced No Significant Hazards Considerations (NSHCs) are contained in enclosure 4. Only technical changes are discussed; administrative changes (i.e., format, presentation, and editorial changes) made to conform to NUREG 1431 Revision 1 are not discussed. For enclosures 3A, 3B, 4, 6A and 6B, text in brackets [ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plaat specific information in that location. CHANGE NUMBER NSHC DESCRIPTION 1 01 M In accordancc with NUREC-1431 Rcv 1, the

• indicated 0 3.5.1-1 tank icvel is added to clarify the volucc. In eddit4ent The plant specific parameters for M accumulator tank icvel and tank pressure ere.is revised to provide limits that can be dire petl [ 0-3.5.1 3 1 confirmed using control room instruments /This'is a /

I ~ use it' reduces the allowable [mo3FestrictivechangebTe pressure: range;1or the: accumulator 7 cover _ pressure, LThere are"actually;two; changes being.made.1The;first change corrects an. existing error'in the_ units lin the CTS, :The existing CTS pressures.(603land 693)(are' psia numbers lnot psig aslindicatedlinlthe.CTSE;Thel correct lpsig' numbers would.be 588)andl678.LThis results.inzthe higher CTS l pressure being non conservative.1The existing i impl ementation ' procedure . for ;thel accumul ator; pres sure _had the correct. pressure range. . The:second change applies j instrument error to the analysis l pressures resulting in new pressure range 1623;psigLto 644lpsig; iThis change allows the use.of controlLroom level: indication;to_ perform the surveillance { These changes arciadministrativc as therc arc no technical differences in thcsc numbcrs. Qhcrc arc no technical changes to thc analyscs. 1 02 A Consistent with NUREC-1431 Rev 1. thc ncainal tank 0 3.5.1 1 valuac increase af-101 galicas is clarified to bc it of tank levci which is the paramcter monitored by the I eperator This changc is administrative as there arc no technical differences in the two numbers. Not used. 1 03 A Replaces reference to the " pressurizer pressure" with a reference to the *RCS pressure". ACTIONS a and b require reducing pressurizer pressure to less than 1000 psig. l However, pressurizer pressure instrumentation does not have the range to read that pressure. Consequently RCS pressure instrumentation is used. For the purposes of this LCO, the use of RCS pressure is equivalent. , 1 CPSES Description of Changes to CTS 3N.S 1 7/29/98 l - - - - - . ----_______J

l ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: 03.5.1-6 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 1-07 LG CTS 4.5.1.1.b (DC, CA, WC) CTS 4.5.1.b (CP) ITS SR 3.5.1.4 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. ) Comment: Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG Response: DOC 1-07 LG has been revised to provide additional justification for the proposed change by adding the following information:

 "The RWST has its own LCO and SRs to verify OPERABILITY and cross-references to other               I specifications are generally inconsistent with the ITS format and are not required to impose OPERABILITY on the referenced equipment. The RWST boron concentration is maintained                 I between [2400] ppm and [2600] ppm which is higher than the minimum boron concentration required to be maintained in the accumulators. If there were reason to doubt the RWST boron concentration, ITS 3.5.4 Condition A would be entered with its 8 hour Completion Time. In addition, ITS SR 3.5.4.3 verifies the boron concentration of the RWST every 7 days. Therefore, it is unlikely that the boron concentration being added to the accumulators would be below

[2400] ppm. Additionally, plant procedures implementing the SR 3.5.1.4 Bases specify that if the RWST has been diluted since its last boron concentration sample per SR 3.5.4.3, the boron concentration in the accumulators must be verified within 6 hours after adding [101] gallons or more to the accumulators from the RWST. The moving of this das to the ITS Bases maintains consistency with NUREG-1431; retention of this detailin the ITS is not necessary to adequately protect the health and safety of the public. Details for performing surveillance requirements are more appropriately specified in the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the plant conditions appropriate to perform a surveillance test is  ! an issue for procedures and scheduling and has been previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is consistent with the vast majority of other surveillance requirements that do not dictate plant conditions for surveillance. Any change to this detail will be made in accordance with the j Bases Control Prcgram described in ITS Section 5.5.14." ' ATTACHED PAGES: i i Encl 3A 2 j \  : l

CHANGC NUMBER HSfE DESCRIPTION 1 07 LG The SR currently requires a 6 hour surveillance if the makeup source is the RWST and the RWST has been di'uted since verifying its boron concentratica per The RWS7 LCO. The proposed change would move the statement "and the RWST has not been diluted since verifying 3.5.1 6

                                                       ...   " to the ITS SR 3.5.1.4 Bases. This level of         j detail is at includad in the ISTS and is consistent          I wi            e kind of informaEo6conte 4ned in the Bases Jhe lhasMtslown LC01and SRs;toi;verifyiOPERABILIThand crosstreferencesitoiotherjspectf1cationsIareigenerally inconsistent with:the:ITS: format'andrereinot:: required:to impose;0PERABIt.ITEonithe;referencodiaquipment;1The;RWSI baroniconcentration;Is;.paintained;between;[2400);ppsland

[2600Eppe;whichMsihigher;thanthe:minimumboron ], concentration 1 required;to;be: maintained;jnithe accumul atorsEIfj thereiwereireasonito;dnubt1the;RWST; boron concentrationgITS;3;5,4;Conditim!A would:be; entered;with its;81 hour; Completion TimeEIn3dditionCITSJSR;3',5c4l3 ) yerjfjesithe: boron concentrationlof1the RWSTieVery;71 days! l Thereforeuttiis:unlikely;that;the boron concentration I beingladdeditoltheaccumulatorswould;be:belowi[2409Ippe;  ! AdditionallyHplant! procedures 11 implementing;the;SR 3;5?l;# Basestspecifylthat;1f;the_RWSTlhasibeenidilutedisince2its last; boron; concentration;; sample; pet;SR;3;E4;3,%thejboron concentration d nithe;acqueul ators laustibe;veti fied ;within j 6ihourst4fterladdingd[101][gallopsion;aoresto;the accumul atorsJfrom1the;RWSTcThe ' noving;of;thi s;detaillto { the ITS; Bases maintains 1consistencyMithiNUREGi14311 retentioniof;thisidetailgin the:ITS11slnot:necessary to j i adequately;protectj the l health tand! safety;ofS the;publ ici ! Details;foryperformingisurveillanceyequirements:are;more appropriately 1specifjedligithelplantiprocedures; required ! by2ITS;514;1:anditheNTSIBases Econtrollof;the? p] ant i i conditions]appropriatesto;perforniaisutveillancestesty1s i antis.suejfotiproceduresMnd;schedulingtand;has;;been previouslyideternined: byPIRC;to~ be; unnecessary;as; a;TS restrictionEAsiindicatedlin;Generjc;Lettet:91304? i a))owing thisflicenseefcolltroEis;consistentiwithithelvast ! majotityiof;otherzsurve11]ance1requirementsithat:doMot dictate' plant; conditions 3focisurve111ancesRAny change;to thisidetaillwjll;be;sadelinjaccordance:withqhe; Bases Control" Program:describedlin1ITS;Section 5' 1 08 A NotapplicabietoCPSES. See Conversion Comparison Table (enclosure 3B). 2 01 LG Consistent with NUREG 1431 Rev 1. the LC0 and ACTION a are revised to replace the word " subsystem" with the word

                                                       " train" and the descriptive information in the LC0 is moved to the BASES. The; proposed; change;is                   0 3.5.2 1 consi stent; withiNUMARCi93103U " Writer,is1Guideffor;the CPSES Description of Changes to CTS 3N.5                 2                       7/29/98
                                                                                                                                )

l l I I ADDITIONAL INFORMATION COVER SHEET  ! ADDITIONAL INFORMATION NO: 03.5.2-1 APPLICABILITY: C 4, CP, DC, WC REQUEST: DOC 2-01 LG CTS 3.5.2 LCO ITS 3.5.2 LCO The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additionaljustification as to why this detail is not necessary in the ITS. FLOG Response: DOC 2-01-LG has been rev;0ed to provide additional justification for the proposed change by adding the following information:

             "The proposed change is consistent with NUMARC 93-03,' Writer's Guide for the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the LCO describes as simply as possible the lowest functional capability of the system and relegates the details of what constitutes an OPERABLE system to the Bases. Therefore, the details of what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY."

ATTACHED PAGES: Enci 3A 2 and 2a

CHANGE HutBE8 NSHC DESCRIPTION 1 07- LG The SR currently requires a 6 hour surveillance if the makeup source is the RWST and the RWST has been diluted since verifying its Mrca concentration per The RWST-LCO. The proposed change would move the statement "and the RWST has not been diluted since verifying 0 3.5.1 6

                                ... " to the ITS SR 3.5.1.4 Bases. This level of detail is not included in the ISTS and is consistent with the kind of information contained in the Bases. The PWSTlhasMts;oun:LCMatSRsitoiverifyLOPERABILIUland grossireferencesitol;otherlspecjficationsIare; general]y inconsistent 1with;the3TS2formatlandlareinot; requited;to impose 0PERABRITEon:theireferencedlequipment0The;RWST boron _ concentration;js; maintained;between![2400Kppeand  z

[26co) ppu yhich11sihjgherdthan;$he:stgim m; boron concentration; requited;tolbes samtpinedMnithe accumulatorsgIf;there:were2easonito_ doubt;the';.RWST borpn concentration;IITS 315;tConditiammmid be; entered;with jts;aihour;Complotton;TimerzIn" addition 11TsiSR13.;5.4(3 yetitiesitherboronconcentration"ofttheiRWSTieveryj7;daysi Thereforedtilslunlikelylthat;theLboron concentratjan bejpgladdedito;theleccumulatorsyould;belbelow;[24co);:ppet , Addjtjenallygplantiprocedures 1mplementingithelSRi3;5;1M Basestspecjfy3hatMfithe:RWSTibasibeenid11uted;since;1ts { last* boron concentration;saepleiperjSRI3;.5L4?3athe boron concentrationjip;theiaccumulators;sustibelverified31 thin 61houtslafterladdingl[101): gal]onstor;noresto3he accumulators 1from1the; RWSI;'Jhe Ming ;ofathis ; detail Lto ' thei!TS:Basesina19tainsiconsistency:With1NUREGt1431l retention;ofithis1detalliin3he1ITSlis1Not1necessary?to adequate 1rprotect3heihealthiand s pafetyiofithe! public; Deta1] sfforiperforminglsurve11)ance fequirementsiare : more approptiatelt specifjed tj n d thel plantiprocedufpsifeguiced hylITS15M;1:andithe ITSiBasesuControEofithe; plant conditionslappropttateltof perform;aIsurve1]1anceitestlis i ands _suelfot: procedures;and,scheduljng:and;hasibeen previous]yldetermined;bylNRCitolbel unnecessary.as2a iTS restrictionZAs3 dicatedI14GeneticjtettetJ1:04? allowing;thismase contro12js:consistentiwithithelvast majoritKof;other; surveillance requirementsithat;dolpot dictate; planticonditions;fotisutve111ences Tl Any;changelto l this;deta11Ew1Ulbeiandednlaccordance:withithetaase_s l pontrol'frogram described.31nJITS;Section:5;5:14] 1 08 A Not applicable to CPSES. See Conversion Comparison Table (enclosure 3B). 2 01 LG Consistent with NUREG 1431 Rev 1, the LC0 and ACTION a are revised to replace the word " subsystem" with the word

                               " train" and the de                            e Irn ic  '

moved +^

• D^[scrinHa inw-e da The:proposedichangeljs nsi stentiwith' NUMARCi93103 Oriter's; Guide; forithe CPSES Description of Changes to CTS 3N.5 2 7/29/98 i

CHANGE NUMBER HSliC DESCRIPTION Restructured;Jechnica11 specifications Land;the

o. m .1 ph11osophyiofiNUREG11431M n Aictthe1LCO:describeslas simplylas; possjble;thellowest; functional 1capabil.ityzof theisystem and[telegatesitheJdeta1]slofmwhat; constitutes an;0PERABLE[systenitolthe BasesEThereforeEthe;deta11.s l of;what constitutes 1an;0PERABLEisubsysteel(train)isuchtas required!pumpscheat;exchangers1andif]ow;pathsgaremore l appropriately; discussed:1nithe; bases 1than;10the;LCO2 These[detai]slareinotinecessaryLtoiensure;ECCS;0PERABILITY

!' otithatithelECCSicaniperfore11tslintended:safetyLfunctions Therefore2the;proposedichangeisoves tolthe Basesideta11s z thatiate; pot.s necessaryito1 provide;operationa13safetyiwh11e retaining;,1ttheitschnicallspecif1catjonsithe; bas.ic l'equiMs;forlmaintaining,0PERABILITYO Whereas there is no technical change associated with the replacement of i the term " subsystem," " train" better describes that all parts of the required system ( e.g., piping, instruments, cont @ etc..) must be operable to support the required safety functions. 2 02 LS-1 Consistent with NUREG 1431 Rev 1, a note with respect to RCS pressure isolation valve testing is added to the LCO. ' Plant design requires closure of certain valves in SI injection paths to perform PIV testing. Isolation of the injection paths in H0DE 3 is currently prohibited as it 1 would constitute entering TS 3.0.3 since both SI trains 4 would be made administratively inoperable. In actuality, the flow paths are readily restorable from the control room and a spurious single active failure is not likely in the short term (2 hours). The new note will allow q closing these valves without declaring either SI train i inoperable. This change is consistent with traveler TSTF 153. 2 03 LS 2 This change revises Action a to allow for increased flexibility in plant operations under circumstances where components in opposite trains are inoperable, but at least i 100% of the ECCS flow equivalent to a single OPERABLE ECCS I train is available. Due to the design of the ECCS l subsystems, the inoperable condition of one or more ' components in each train does not necessarily render the I ECCS inoperable for performing its safety function. The allowed outage time of 72 hours is unchanged: but, it is to be contingent on being capable of providing 100% of the l ECCS flow equivalent to a single operable ECCS train. This , change is consistent with NUREG 1431 Rev 1. l 2 04 TR 2 Consistent with NUREG 1431 Rev 1 the requirement to submit a Special Report within 90 days of an ECCS I CPSES Description of Changes to CTS 3N.5 2a 7/1988

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.2-2 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 2-09 LG CTS 4.5.2.c The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this surveillance is not necessary in the ITS. FLOG Response: DOC 2-09 LG has been revised to provide additional justification for the proposed change by adding the following information:

                                                                                                                                  " CTS SR 4.5.2.c requires a visual inspection to verify that no loose debris is present in
                                                                 ,                                                                the containment which could be transported to the containment sump and cause restriction to the pump suction during LOCA conditions at the frequency specified. This ensures that during the process of performing maintenance or other work inside containment that debris is appropriately discarded. Existing procedures restrict containment entries and assure accountability of items entering containment such that they are removed at the completion of the containment entry. ITS SR 3.5.2.8 continues to require a visual inspection every 18 months on each of the ECCS train containment sump suction inlets to ensure that the sump suction inlet is not restricted by debris.

Therefore, this detail is not required to be in the technical specifications and moving this requirement maintains consistency with NUREG-1431." ATTACHED PAGES: Enci 3A 3 and 3a I 1

  - _ - _ _ _ _ _ _ - - - _ _ _ - - _ _ _ _ _ _ _ - _ _ - - _ .                                                                                     _                                                                          l

CHANGE NUMBER HSBC DESCRIPTION actuation and injection event is deleted. This change is acceptable because the requirement to submit a report is sufficiently addressed by the reporting requirements contained in 10CFR50.73. 2 05 LS 3 This change revises the LCO applicability note to allow operation in MODE 3 pursuant to LC0 3.5.3 until "all" cold legs exceed the RCS temperature setpoint in lieu of "one or more." The previous allowance was [for 4 hours or until "one or more" cold legs exceed 375"F. whichever comes first]. The 375'F is a nominal temperature selected to give time to restore tha pump operability without delaying startup. The four hour limit is unchanged. Changing "one or more" to "all" is still bounded by the 4 hour limit. This change is consistent with NUREG 1431 Re/1. 2 06 Not used . 2 07 A Consistent with NUREG 1431 Rev 1. this change revises the surveillance to make it clear that " listed" valve position is the concern and not indicated position in the control room. The surveillance can be satisfied using indicated position in the control room but may also be satisfied using local observation. This is an administrative change since the surveillance acceptance criteria are not changed. 2-08 A The accumulator discharge valves and their note are functionally part of the ECCS accumulator subsystem covered by improved TS 3.5.1 and are surveilled by SR 3.5.1.1 (although the specific valve numbers are no longer included in the improved TS). They are not part of the LC0 , function in the ECCS LCO. This is an administrative change to clarify the location of this requirement. i 2 09 LG The visual inspection surveillance performed when establishing containment integrity is moved to a licensee controlled document. CTS?SRJ4:.5.2;cirequires 0 3.5.2 2 ; alvisuallinspection).toivetifyLthat2nojooseidebrisiis Present2initheicontainment:which coul.dibeltransported tolthe1 containment; sump;and cause; restriction;tolthelpump suctioniduringLOCAlconditionslat;theifrequencyJspecifiedB Thislensures;that:during;the:processloffperforming i maintenance [orjothettworkiinside;containmentithat" debris islappropCiately(discarded.R Existing procedures 1 restrict contajnsentientties3 nd: assure accountability;ofditems enteringicontainment(suchithatttheyfare: removed:at;the compl.etionjofithel containment entry O ITS SRL3.5'.2.8 continuesitoir_equireya(v1sualjinspection every 18; months onleach;ofithe;ECCS t train:containmentisumpisuctionlinlets CPSES Description of Changes to CTS 3N.S 3 7/29/98

CHANGE 1 ! NUMBER R$liC DESCRIPTION I o ensureithat thetsump:suctionfinlet~1sinot 0 3.F 2 2 l restricted:by debris C Therefore,ithisideta1111sinot i required;to be:in theitechnicallspecificationsl.and l moving this requirement maintainslconsistencylwith;NUREG. 1431; Memg-this typc of rcquirc=r.t is corsistcr,t with o- m. .m .. 1 { 2 10 A Consistent with NUREG 1431 Rev 1, the current TS SR for l verifying interlock action of the RHR system is moved to I improved TS SR 3.4.14.2. l l 2 11 TR 1 Consistent with NUREG-1431 Rev 1, the ECCS pump and valve actuation SR is changed to allow the use of an actual l l signal, if and when one occurs, to satisfy surveillance I requirements. The specific signals used to actuate the pumps and valves have been moved to the Bases. l

                                                                 .In;; 1severa1     Specifications %throughout EtheZCTS?

0 3.5.2 3 OPERABILITYiofJ;.certain Zequipment sisidemonstratediby ensurj ng'sthat D;the Lequipment 2 performs 5;1tsnsafety functj on :Lupont receiptiofia ? simul ated; test Esignal aThe

                                                                 .intentiof'ajsimulated*3 signal wasitp;be;able;to;peCfors;the requiredLtesting;.without:the; occurrences (or;Without; causing) an actualesigna1Egeneratingtevent; LHoweyersthe unintended e_ffectiwa sito 1 requi re ~;the; performance; ofsthelsurve111 ance (u si ng L antest ;lsi gnal );;even 21 f n anHactua.12signalf had previouslyl;vetified theioperationfofithelequipmentMThis change;all ows ; credit ; to : be: taken iforf actualievents? wheqlthe required equipmentjactuates.;successfully; While Ethe1 occurrence 1:. ofievents; that s causelectuationgof accident mitigation; equipment is1 undesirable ~,1the actuation                                           1 ofMmitigation1equipmentnon an;actua12 signa 1EisCa"better                                                 l demonstration [offits;0PERABILITYithan;an1 actuation using;a                                              I testE; signal l. , LThus; . Lthef change Cdoes; notEreducerthe re.11abilitylof 2.thelequipmentEtested:J 1TheEchange "also improvesT pl ant 1 safety; by; reducing;the: amount [of[timel.the equipment?.is;takenLoutiof service;for testingE andithereby increasing jtsiavailab.ility"duringiang actualfeventt and. by

reducing ? the.; wear of ;the) equipment ; caused { by; unnecessary l testing; 1

CPSES Description of Changes to CTS 3N.5 4 7/19/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.2-3 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 2-11 TR-1 CTS 4.5.2.e ITS SR 3.5.2.5 & SR 3.5.2.6 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it le consistent with the STS. Comrnent: The NSHC for this change appears to provide the needed justification. Therefore, please incorporate the information contained in the NSHC into the subject DOC. FLOG Response: The CTS requires the use of a test signal for initiation of valid tests. The unintentional result was to require the performance of the verification even if an actual signal has already verified proper operation of equipment. TR-1 allows either an actual or test signal. DOC 2-11 TR-1 has been revised to provide additional discussion to allow to the use of an actual signal to meet this surveillance requirement. ATTACHED PAGES: Encl 3A 3a l 4 I l l l l . i l _ - ____ _______ _ _ ~

CHANGE NUMBER N2iG DESCRIPTION 3.5;E8; continues;tottaquire;aivisuallinspection o.3.5.2 2 SveryJ181monthslogteach;ofithe!ECCSitrain; containment i susp;suctionMnlets;tolensurelthat;thelsump1 suction L jnletliis;M restrj.cted;by;debrjsF2Ihereforeilthisidetail 1 s[notitequiteditolbeljn), thel $echnica]1specif1cationst and soying1this; requirement lmaintainsicons1stency;jvith;NUREGi 1431 ",cvini; this typc ;f requir;xat is censistent with "UREC-1431 Rcv 1. 2 10 A Consistent with NUREG-1431 Rev 1, the current TS SR for verifying interlock action of the RHR system is moved to improved TS SR 3.4.14.2. 2 11 TR 1 Consistent with NUREG 1431 Rev 1, the ECCS pump and valve actuation SR is changed to allow the use of an actual signal, if and when one occurs, to satisfy surveillance requirements. The specific signals used to actuate the pumps and valves have been moved to the Bases. s -

                                                                                                        -                                                                     l nZ severalfaspecifjcations;2 throughout gg the];1 CTS; o.3'5'2 3 OPERASILITYIofIcarta192equipmentIlsidemonstratediby ens _urjng3that titheXequipmentZ perfores ajts Esafety
                                             ,                                                    functioniuponiteceiptXof2 a2simulateditestisignalgi?Ihe                     !

I fntent:ofXsimulatedrsjonahwasito:belable:to; perform;the j requireditestingMthout1theioccurrencel(egthouticausir;gl j an;actua11 signa 1Jgeneratjng; event E However gthelugintended i effect1wasitoirequireithe;performancelofEthe1 surveillance (usjpggeEtestKsignal)2evenMifianXactualXsignalghed prey 1ouslyiverifiedf,thef operationlof1thelequj pment GJhi s changelallows; credit 1to'be taken;fot;actualievents;whenithe requiredlequipmentractuates;succes_sf4]1yj j Wh11entheroccurrenceto(KeventssthatIcauseyactuationiof I a_ccidentimitigation :equipmentM slundesirableRthelectuation off aitigationiaquipment ronranIactuaEsignal uj sla5 better demonstrationiof[its10PEABILITYithan:an;a.ctuationfusingEa

                            !                                                                     testIsigna17,TEtJhus;Ethe 2changeldoesEnotgreduceEthe relj ab1]jtyXofathelequipment$ tested @IheHchangei ai so japrovesip]antIsafetys byIreducing3the:amountiofitimelthe equipment is;takenioutiofiservicelforJ.testinggand;thereby increasinglitslayall ability;during f aniactuallevent Cand ; by reducing ;the twearlofithe p equipmentReausediby1 unnecessary testing {

l j ~__ CPSES Description of Changes to CTS 3M.S 3a 7/29/98 i

l ! ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.2-4 APPLICABILITY: CA, CP, DC, WC 1 REQUEST: DOC 2-12 LG I CTS 4.5.2.f ITS SR 3.5.2.4 l The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail is not l necessary in the ITS. FLOG Response: DOC 2-12-LG has been revised to provide additional justification for the proposed change by adding the following information: 1 l

                     "lTS SR 3.5.2.4 retains the SR requirement and references the Inservice Testing Program (IST),

discussed in ITS 5.5.8, for the surveillance Frequency. The specific SR acceptance criteria for the J pumps have been moved to the ITS SR 3.5.2.4 Bases. Although this may make the ECCS pump performance testing more flexible in the future, only with regard to licensee control over the numerical values of the acceptance criteria, this testing must continue to conform to the IST program requirements. Revisions to the acceptance criteria will have to meet the requirements of the Bases Control Program discussed in ITS 5.5.14. Details for performing surveillance requirements are more appropriately specified in the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the acceptance criteria for a surveillance test is an issue for the IST procedures and has been previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is consistent with the vast majority of other surveillance requirements that do not dictate plant conditions for surveillance." ATTACHED PAGES: i Enci 3A 4 I i t i

CHANGE NUMBER tiSBC DESCRIPTIQt{ 2 12 LG The ECCS pump performance is revised to be consistent with NUREG 1431 Rev 1. The test method and specific data required to verify pump performance is moved to licensee controlled documents. Specification 4.0.5 no longer exists in the improved TS: however, the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 of the improved TS. The IST Program is referenced directly for M the freque.ncy of testina3!KSR;3,5.2;Lreta1nsithe;SR o 3.5.2 4 1 requirement;andfeferencesLthe:Inservi_ce Testingfrogram (IST)% discussed lin E ITS 15]5;8lifot;thensurve111ance FrequencysJhelspecific;SR acceptance criteria forithel pumps have;been[ moved;toltheilTS SR1315;2'4 Bases;ZAlthough;this sayimakelthe ECCS pump; performance; testing.more;f]exiblelin the 2 future Con]$ withd regard;tolicensee;controMoverithe numerica1Xyel@sXoEtheJacceptance criteriai;lithisitesting j mustf;continueltoiconformitolthellSTiprogramirequirements? I Rey 1 sions;to ; thel acgeptance:griteria y11E havelto; meet;the requirements;ofitheiBases;ControRProgramidiscussed11n1ITS 5.5114;; Details;foriperforming: surveillance;requirementsjare morezapproprj ate]yZspecifiedlJ i n.E theE pl ant E procedures requi ted; byEITS;; 5 ;4:1 E and EtheIITS;Ba sesU Contro11 ofithe agceptance icriterf alforlal survc1]1 ance ;testlis; anf j ssue Joc the;IST procedures;and;haslbeen,previously determined by)NRQ tolbeiunnecessary tasZalTSy restrictiontAsEindicatedEin Generic ;LetterL 91h04Ea110 wing ithi sD icenseef control " 1 s consistent [withat.heEyastimajoritylofr thetEsurve111ance o requi rements Ethatido E notidi ctate s pl ant ;conditionsTfor surveillance; 2 13 TR-3 The CTS allowance. which permits the ECCS throttle valves to be declared OPERABLE without verifying ECCS throttle valve stop position for 4 hours following valve stroke testing or maintenance is deleted from the current TS consistent with NUREG-1431 Rev 1. The ECCS throttle valves are manual valves and plant procedures governing post maintenance test requirements specify verification of correct throttle position prior to declaring the valves OPERABLE. Explicit post maintenance TS surveillance requirements have been deleted because these requirements are adequately addressed by administrative post-maintenance programs. 2 14 A The note providing a one time extension of surveillance intervals is administrative 1y deleted since it is no longer applicable. 2 15 LG The surveillance requirement for the flow balance test following ECCS modifications is moved to a license controlled document. This1 requirement:1s.not included 0 3.5.2 5 CPSES Description of Changes to CTS 3N.S 4 7/29/98 l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: 03.5.2-5 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 2-15 LG CTS 4.5.2.h The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this surveillance is not necessary in the ITS. FLOG Response: DOC 2-15 LG has been revised to include additional justification as to why this surveillance is not necessary in the ITS. ATTACHED PAGES: Enci 3A 4 and 4a i

                                                                                                        )

I I l 1 _______________ _ _ D

1 l CHANGE NUMBER NSE DESCRIPTION 2 12 LG The ECCS pump performance is revised to be consistent with NUREG 1431 Rev 1. The test method and specific data required to verify pump performance is moved to licensee controlled documents. Specification 4.0.5 no longer exists in the improved TS: however, the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 of the improved TS. The IST Program is referenced directly for the frequency of testing. ITS1SRf3;5]2;4EretajnsithelSR o 3.5.2 4 requirement and:referencesithe:Inservicelestingfrogrom (IST)rdiscussedI1n;ITS.55:5;s;Ifor.ithe surveillance frequencyaTherspecific;Stacceptance;critetiafforithe; pumps have: beetmoved3o3hei!TS;5R23;5;2:4;BasesgA]though:this pay;makeithe;ECCS:pumpiperformanceitesting;noreiflexibleMn the3futuremlylwithtregard2tomjcensee:controroverathe numer,1 calf;valuestofitheracceptanceterjterja t,this1 testing musticontinuelto; conform;toltheJSTEprogram: requirements Reyjsionsitolthe; acceptance 1criter. jay 1]Ehaveltoimeetithe requjtementsIofithe1 Bases 1ControliProgras? discussed;1n;ITS E5;149 eta 11s:for:perforeing;survet] lance: requirements are poterapproptiately1specified Ti1REthe 2 pipntR ptocedures requitedibylIT3gs:411randitheEITS; Bases &Contro12cfs:t:he acceptance;ctitatiafforlaisurye111 anceitestli sian if ssuelfor theilST; procedures;andj hasibeen:previously; determined _ byJRC to2belunnece_ssarylassa ETSXtestrictionMsiindicated31g Generjc;LetterH91104Ha]1owingiithi.siljcenseelcontrollis consistantswithltheEyastraajorityIoflothergsurvejilance requirements 1thatZdolpot E; dictate!!p] ant *Lconditionsj;for

                               .sutye1]1ancest 2-13          TR 3           The CTS allowance, which permits the ECCS throttle valves to be declared OPERABLE without verifying ECCS throttle valve stop position for 4 hours following valve stroke testing or maintenance is deleted from the current TS consistent with NUREG 1431 Rev 1. The ECCS throttle valves are manual valves and      plant procedures governing post-maintenance test requirements specify verification of correct throttle position prior to declaring the valves OPERABLE. Explicit post maintenance TS surveillance requirements have been deleted because these requirements are adequately addressed by administrative post maintenance programs.

2 14 A The note providing a one time extension of surveillance intervals is administrative 1y deleted since it is no longer applicable. 2 15 LG The surveillance requirement for the flow balance test following ECCS modifications is moved to a licens3 controlled document. . is requirementiistnotxincluded 0 3.5.2 5 CPSES Description of Changes to CTS 3N.S 4 ~,':" "'

• L ,

J ( CHANGE I NUMBER HSRC DESCRIPTIQH nLNUREG:14311 lP1 ant proceduresigoverning1the

                                                                                                     'D 0-3.5.2-5 restoration l,0f; equipment after maintenancetspecify the~ requirements;forideterminingithe;; appropriate postimaintenancettesti.ngnllAny time the;0perability,of a system oricomponent;has beenjaffected1by repair;                                ,

maintenance Cor; rep 1acementiofia componentt post 5 l maintenancEtesting;is required;to;6emonstrate_0perability I oflthe1 system or; component.i As such;1the requirement to perform a" flow balance; test lafter modifications:thatf alter l ECCS subsysteufflowicharacteristics 1s'notl: required.:tolbe in;the;TSito provi_de adequate protection lof;the pub 1.1.c healthiandlsafetyMJhisirequirementlhas beenLaoved,to the [F]SAR;(foruCallawaylJDiabloiCanyon;;and WolfjCreektor TRMJComanchePeak)f.lThese: licensee 1 controlled: documents containing;the; moved 1requirementsWill;belmai.ntained using j the' provisions;ofl10;CFR150.59 iThereforen the; descriptive l information,that has;been, moved l continues;to be maintain in{an appropriately lcontrgil.ed manner.: _ 2 16 LG The specific means by which the ECCS piping is assured to be q full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with the kind of information contained in the Bases. The; requirements of ITS 1.00'3.5;2cand the! associated, Surveillance { l

                                                      ~

au24 Req'uirements(e.gf,:ITS:3'5;2;3)Jareladequatefto ensure;the,ECCSLare maintainedj0PERABLEL JAs;afresul.t'i the methods of, performing Surveillance lare.not necessary to ensure the ECCSican: perform,their21ntended safety; function anditheidetailsiareinot:requiredLto be11n the TSlto provide adequate protection,0f: the' public health and safety! 1TheLITS Bases;containing,the moved requirements _will[be maintained using;the: provisions ofc10 CFR(50 59lJasirequired by. Bases Control' Program.de_ scribed inLITS:SectionL5.5.14C Thereforet 4 theidescriptive;information;that:hastbeen moved continues to ' I beTmaintained;in anzappropriately l controlled mannerp I 2 17 A Adds the phrase 'that is not locked, sealed, or otherwise secured in position' with regard to which valves require actuation testing. This change is merely a clarification. , Valves that are secured in place, are secured in the position i required to meet their safety function. The actuation testing ensures that valves can move to the position that meets their safety function. If the valves are secured in the position that meets their safety function, no testing is l necessary. 2 18 LG The requirement for venting the ECCS pump casing and accessible discharge piping high points following any maintenance or operations activity which drains portions of the system is moved out of the TS consistent with NUREG 1431 Rev 1. Plant procedures governing the CPSES Description of Changes to CTS 3N.5 Ja 7/29/98 l ._______________________A

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.2-6 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 2-16 LG CTS 4.5.2.1 ITS SR 3.5.2.3 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG Response: DOC 2-16 LG has been revised to include additionaljustification as to why this detail is not necessary in the ITS. ATTACHED PAGES: J Enct 3A 4a ( l l l I ' i 1

CHANGE NUMBER RSE DESCRIPT10N jn;NUREG:143171 Plant? procedures; governing!the 0 3*5'2-5 restoration ;ofiequipmentfafterlmaintenanceispeci fy thelrequirementsifor1 determining;the;appropr.iate posthmaintenanceitestingOlAny;timejtheloperability.of;a system;oricomponentihaslbeenfaffacted;byirepairl maintenancegorireplacementofiafcomponent,; posts maintenanceitestingli s! requi reditoldemonstrate ;0perab111ty of;the? system ~for(component;TAsJsuchsthe; requirement;to perform;a ; flo(balance itestsafteC, modi.fications tthat L al tet ECCS: subsystem;flowicharacteristics]1s;notirequiredito be in;the1IS;to; provide: adequate; protection:of;thepublic health andisafetyEThi_sirequirement;hasibeenjeoved;toithe [F] SARI (fot; Cal) aways Diablo; Canyon lf a.ndf Wol f; Creek); or IRMJComanchePeakE.Theselicensee;; controlled; documents containing ,the: moved ; requirementsiw11Ebe ' maintained :using the: provi sionsioffl0;CFR; 50; 59 ETherefore atheidescr:1ptive information(that has;been' s moved; continues;to be maintained in anzapproprjately1 controlled 1 manner! 2 16 LG The specific means by which the ECCS piping is assured to be I full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with the kind of information contained in the Base f The; req Mus y lIA1.CD;3.5:2:anditheJssocatedSurveillance Requirements:(e;gt,1ITS5375;2:3)[aretadequatelto Io3.s.26 I ensurejtheiECCS1reimaintainediOPERABLEOfAs;airesult? the methodslof; performing:Surve111ancesiaresnotinecessaryito ensure;the.ECCSicanfperforpitheidintendedlsafetyifunction anditheidetailslareinot requiredito;belin;the!TSito provide adequateprotectionl;of;the! pub 11c;healthiandisafet/C.ThelITS Bases 1conta.ining;the moved l requirement.s;will;be; maintained using;the provisjonsiofi101CFR150;59,Eas1 required by Bases ControliProgramideset1 bed iin lITS ;Section15l;;5 :14;;Therefore " the;descr;lptivelinformationjthatihas;been;movedLeontinues;t be'saintainedLinJan2 appropriately; controlled ma 2 17 A Adds the phrase 'that is not locked, sealed, or otherwise secured in position

• with regard to which valves require l

actuation testing. This change is merely a clarification. Valves that are secured in place, are secured in the position required to meet their safety function. The actuation i testing ensures that valves can move to the position that meets their safety function. If the valves are secured in the position that meets their safety function, no testing is necessary. 2 18 LG The requirement for venting the ECCS pump casing and accessible discharge piping high points following any maintenance or operations activity which drains portions of the system is moved out of the TS consistent with NUREG 1431 Rev 1. Plant procedures governing the CPSES Description of Changes to CTS 3M.S ha 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.2-7 APPLICABILITY: CP REQUEST: DOC 2-17 A CTS 4.5.2.e.1 ITS SR 3.5.2.5 The Conversion Comparison Table identifies this change as applicable to Comanche Peak and the change is included in the ITS; however, the CTS markup does not include this change. Comment: Please revise the CTS 3/4.5.2 markup to reflect this change. FLOG Response: The CPSES CTS markup has been revised to include this change. ATTACHED PAGES: Enct 2 3/4 5-5 ___________-__________________---J

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS

d. At least once per 18 months by:

l

1) Verifying interlock action of the RHR system from the Reactor **** M I Coolant Systei to ensure that with a simulated or actual Reactor I I

Co31 ant System pressure signal greater than or equal to 442 psig the interlocks prevent the valves from being opened.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and -

that the sump components (trash racks, screens, etc.) show no j evidence of structural distress or abnormal corrosion. 1

e. At least once per 18 months *, by: #216A 9 '

1) Verifying that ach automatic valve in the flow nat that 2 17 A isinotllockedssea ec or upieFw1se,securedlin' posit 10 0 3.5.2 7 )

actuai.m to i W correct position on 5siscy Injcction actual , o Esimulated actuation test signals, and 12-u'm q

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection an; actual;otsimulated b 211 m actuation test signal:

a) Centrifugal charging pumps, b) Safety injection pumps, and c) RHR pumps.

f. By verifying that coch of the fc11cwing pumps develops the indicated L212m differential pressurc on recirculation ficw when tested pur:usat to Specification 4.0.S; 1

1) Ccatrifugal chrging pump i 2370 psid, 1 1

2) Safety injection pump E 1440 psid, and

3) "J:" pump > 170 p34d-  !

Verify;each ECCS pumplsideyeloped headJat3heitestiflWpointds ) greatercthan ' oriequalitolthe Tequired ; developed;headlinjisecordance; with theInservice;JestingProgram; frequency;

• 4 2-1+Ar, The suneillence test interval is extended to 24 months for Train A, Unit 2, to rc;;in in affect until the c a pletion of the scccad 7dca',i;G c;ets fu 'arit 2.

CPSESMark-up of CTS 3N.S 3M S-S 7/29/98 _ _ _ _ _ _ _ _ - _ _ .__-_______-_-___-_-__-_--__A

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.3-1 APPLICABILITY: CA, CP, DC, WC l REQUEST: DOC 3-01 LG l CTS LCO 3.5.3 l ITS LCO 3.5.3 l The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG Response: DOC 3-01-LG has been revised to provide additional justification for the proposed change by adding the following information:

      "The proposed change is consistent with NUMARC 93-03, " Writer's Guide for the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the LCO describes as simply as possible the lowest functional capability of the system and relegates the details of what constitutes an OPERABLE system to the Bases. Therefore, the details of what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY."

ATTACHED PAGES: Enct 3A 5 ! i l i

CHANGE NUMBER HSBC DESCRIPTION l ' restoration of equipment after maintenance specify the j appropriate post mainter.ance testing. ' 2-19 LG Not applicable to CPSES. See Conversion Comparison Table (enclosure 3B). ( 3 01 LG Consistent with NUREG 1431 Rev 1, the LC0 is revised to replace the word " subsystem" with the word " train" and l the descriptive information in the tco is moved to the BASF1 4 lproposedichangelisfcons WenE yltn M93103EyrJter'siGuidelfor thelRestructured l03.5.31 l Technica11Specificationstand;the philosophy:otNUREG] 143111njwhichlthe;LC0ldescribesias; simp]yfasjpossible_t lowestifunctionalicapabilityofithe; system:andire]egates theldeta11 siof;What; constitutes [an10PERABLEglstem;;tol t l Bases'MThereforetthtdetail.szofiwhaticonsMyJteslan l OPERABLE; subsystem;(train)lsuchlasJrequired:pumpsRheat exchangersland T ow: paths C areinore;approprj ately di.scussed;;1n2the;. bases 3han jnjthe LCOMTheseldeta11s are'not;necessary;to ensure:ECCS"0PERABILITY!;o tthat'the ECCS;can; perform;1ts; intended: safety function d Thereforegthelproposedichangemovesitoithm; Bases; details i that:areinot ;necessary,to; provide;operationalisafety Wh11effetaininglin;theitechnicalsspecifications the;ba irements1forfmaintaining0PERABILITJY wnereas Inere is no tecnnical change associatea witn the replacement of the term " subsystem " train" better describes that all parts of the required system ( e.g., piping, instruments, controls etc..) must be operable to support the required safety functions. 3 02 LS 4 Consistent with NUREG 1431 Rev 1, the low temperature overpressure protection limitation on ECCS pumps and related surveillance are moved to improved TS 3.4.12. The prescriptive wording related to pump operability is changed to wording specifically addressing the pumps' capability to inject into the RCS. This change is less restrictive on the configuration of the centrifugal charging and safety injection pumps but is acceptable because it is consistent with the cold overpressure analysis requirements and still precludes flow to the RCS. Corsistcrt with NUREC 1431 Rcv 1, the LC0 3.5.3 a 0-3.5.3 2 3 03 LS 5 ! Actier, terrir, elegy i; re?i rd r,d th; d;;riptiV; l ir.forr.aticr r,cycd to the BASES. The completion time for l COLD SHUTDOWN due to CCP inoperability is increased from f 20 hours to 24 hours. This time is reasonable based on l operating experience to reach H0DE 5 in an orderly manner, without challenging plant systems or operators, and is consistent with other shutdown action Completion CPSES Description of Changes to CTS 3N.S S 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.3-2 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 3-03 LS-5 CTS 3.5.3 Action a ITS 3.5.3 Actions A & C DOC 3-03 LS-5 discussed two distinct changes. The first change involves movement of the descriptive information to the Bases. The second change is an increase in the completion time to reach Mode 5 from 20 to 24 hours. Comment: The first change, movement of the descriptive information to the Bases, should be separated out and justified as an "LG" change, consistent with other similar changes in this section. The increase in the completion time to reach Mode 5 from 20 to 24 hours is correctly justified as an "LS" change and the justification provided in DOC 3-03 LS-5 is acceptable. FLOG Response: DOC 3-03 LS-5 has been separated into two DOCS (DOC 3-03-LS and DOC 3-13-LG). DOC 3-03 LS-5 has been revised to address only the increase in completion time. DOC 3-03 LS-5 has been enhanced to include additionaljustification. New DOC 3-13 LG has been created to address movement of information to the Bases. ATTACHED PAGES: Enci 2 3/4 5-7 Enci 3A Sa and 6a Enci 3B 4 and 5 t

1

                                                                                                                                    )

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T3 < 350*F l ECCS SUBSYSTEMS t LIMITED CONDITION FOR OPERATION I l 3.5.3.1 As a ;;;inimu;;;. One One ECCS subsystcm, ccaprised of the fclicwing train  ! ** l shall be OPERABLE:N  ! 3.en-u n

                                                                                                                                    ]

s 306 AW

a. Onc 0"E"XLE ccatrifugal charging pu;;;p,'

L

u. A, Anf*M A MI f*

v m, v. m-M

                                       , ,i t,h,
                                              ,  L, m.o. A,. m,L.

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                                                                    .3,,

l c. Onc OPE"XLE "Ji", pu;;;p, and

d. An 0"E"XLE flow path capabic of taking suction from, the refueling water storage tank upon being ;;;;nually realigned and transferring suet 4en to the contain: cat su;;;p during the rccirculatica phase of operation.

6 APPLICABILITY: MODE 4. ACIlM:

a. With no ECCS centrifugalicharging.pumpi(CCP) subsystem OPERABLE 3 13 LG

                                                                                                               ,0 3.5.3 2 because of the inoperability of cither the ccatrifugal charging-pu;rp or the ficw path from, the refueling water storage tank,                                  ~-

restore at least one ECCS CCR subsystem to OPE status within 3 03-LS 1 hour or be in COLD SHUTDOWN within the next 24 hours. 0 3.5.3 2

b. 3 04.LG With no ECCS residual heat" removal!s(RHR) subsystem OPERABLEbe

                                                                          ~

0 3.5.3 3 heat cxchangcr or "J:", pump, immediatelyiinitiatelactionito restore at least one ECCS RHR subsystem to OPERABLE status or maintain the 3 14.A Pacter Ccchnt Sy:; tem-T,,, less than 350 f by use of altcinate 0 3.5.3 3 g heat re;;;; val ;;;ethods.

• A maximum of two charging pumps shall be OPE"XLE capableiofiinjecting;1nto *-M i the;RCS whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F except when Specification 3.4.8.3 is not applicable.

     #       An RHRitrain may,be considered..;_0PERABLE during; alignment and;operationifor                                     .

decaKheat,remoyalRjf; capable;of;.beingl manually 2 realigned totthe;E.CCS1 mode ** H of; operation;f CPSES Mark-up of CTS 3N.S 3M S-7 7/29/98

CHANGE NUMBER list lC DESCRIPTION L' l Times to reach- H0DE 5 from H0DE 4

                                                   ..                                                                                       ~:to:the: stable i tjons;associateo.witn operation 11 nim 00E;4Ethe                                                                                  ,        0 3.5.3 2 probabjlitylofioccurrence:ofia1 Design 18 asis: Accident is31o C h airesult;stheiECCS; operational                                                                                                        {

requirements;are! reduced;Withlonly;oneltrainlofithelECC} CCIE.Subsystes!. required:to:beloper.... abler- _The'r.equired

                                                                                    - - -                                                              ~            - -

actjmI!fsthe;CCPiSubsystes;islinoperable;j s' to; proceed tolcoldIshutdown] s 3 04 LG .V. . d . d b . '..'."b., " .. . . " V" . '

                                                                                              " " . ~ ^'  .. " ~ . ' ' . ,b " . ~ ~^ .~..             V
                                 "^^-'_-'_^.'_b'        _                  _ _ .           ;

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                                                                                                                                               ._                                         0 3.5.3-3
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_ _2. . .v,. V 3,. . 2.._. .s. 2. mV. .. 2.. _ ._ ._~_

                                                                                                                                        . b _.mbV . s d bV, .

least J JAJ.L_ era -_LJ ECCS subsyst;;, is.. revised _ A_ ___A___n nlin _..L_. t; "..in;diately A__ t.f J A L L_AL niin III I b i u b5 Mb L. I VI s bV 3 E d bVI E WII RBHs dWWJJ d bbWI e WW 5 b5 5 UV b 5 4 ITI In pur.ps ___.. ... tu_ or,d_, _. hcot

                                                                   ._ __ ._exch;rgrs unne e                 ir,ep r:ble,              ..L___

it would.u_ _ be unw',,st _ , . . .... .u ,to_ i squ .s b wars y 4 us sb bV yV kV s TVwk &, s sw s w bues vi s u J uvu5aupis I heat r;;.;;ci system is the Thcr;ferc, the appropriate

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_ _Ld_ bV . . I__b d..&VI ___ ___LVVW L VIUb ___IMrrre In non sub:;ystca :nd to car,tinuc the actions until the subsystc; +s A. Anl*M A n f e _ L A. i . _ A1__ bs , y,_.J....~_._.1

                                 .b  __L,,_J   , s.V       bV      V, mnn.mm ., b                 b.        .          m , ., V    Ab .L
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h::t re;;;;l matheds is descriptive infor;;;; tion

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                                  .,___;.. ___L2te                            _;      2_     &La_           ______2.                  L.. .L.          ecce usewwwy yu VIs s u s bwss use bil I d dbbu pus IV WJ                                                        b5fb LVVW
                                  -__a            ,_

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                                                                     . __ u. .nn,re, , , _; ,                                         n drms. . s. . .            b             >V.         V.    -. ..                      ...V     .. .TS,3+5.3,2 Action b ProvideslanlalternatelrequirementMi f;IORicannoti be restored)ito;maintainLE350*Ebyjuse;of?alterpate heat removalimethodsEEThese;deta11 sf are;moveditolthe!ITSiBases G Theseideta11 siareinotipeces.sary, tolensureithe1ECCSlare OPERABLEy2Thelrequirements;ofRITS~ 3 l5;33001:and; Conditions are; adequate;forlensutjpg;that;the;ECCSlare OPERABLE O These                                                                                                      )

detailsiareinot1necessary;tolensureJthe;ECCSican" perform theirljntendedjsafety;functionEAsisuch;?these: details;are notitequireditolbeijnathelTSito provide; adequate lf Vwtjon ofithel public1 health ^andisafetyE1 Moving ;these; detail s maintainsiconsistency,withiNUREG31431 H Any;changeito_these detail s;wi1Ebejmadelin feccordancelwith! 10CFR50,59l and _the BasesiControEProgramidescribedfjp11TSiSectjon;5,5;14y CPSES Description of Changes to CTS 3N.S Sa 7/2988 l

1 CHANGE NLHBER gitE DESCRIPTION actu.ationl;of;the required _ECCS:toimitigate;the 0 3.5 3 3 consequences;ofZaDBAI 3 11 LG Not applicable to CPSES. See Conversion Comparison Table (enclosure 38). 3 12 A Not applicable to CPSES. See Conversion Comparison Table (enclosure 38).

    $f13          LG             CTS 1LCOl3;513JActionlagterminologyli_sirevj sediandj the 0 3'5*3 2 descrjptivelinformation:js:moveditoltheITS1 Bases;g These; detail s;areinot;necessaryj to; ensure;the1ECCSj                          l Required:Actionslarel;metMThe requjrementslofflTS13;5:3lLCO                     I and;Conditons:are adequate;forlensurjng;the!ECCS;areiOPERABLEM These:deta11siareinot necessary;to:ensureytheLECCS;can.; perform theirlintended;sefety1 function &As:suchathesel:detailslare pot:yequj reditolbe:in;the2TSitolprovideladequate; protection;of thelpublic, health"and;safetyMMovingithe.se; details maintains J:ons1 stencyiwithiNUREGf1(3172Anyichange stolthese details;willibeleade;1n:;a_ccordance withi10CFR50;59:and;t Bases;Controlf.P y anidescribed!1nflTS1Sectionl5;5:1 3!14          A              CTSELC0;315;_31Actionibe providesEWithi no1ECCS M o 3.5.3 3 subsystem.0PERABLE7;theioption;tofeither: restore?at least;one;;ECCS;RlRJ subsystenito ,0PERABLEistatus "orito maintain;the;RCSIT y i 350*F;by;uselofdalternatalhcatjremoval.

methods g Condition X ofdITSlLC0;3;5;31 requires;that:withino ECCSRsubsystems,0PERABLELthatiimmediate;actionibe j nitiatedito; restore ;an;ECCS1RlRisubsystenito10PERABLEistatusl2 Whilelthe; CTS 1doesinotispeci fyla; time; frame;toJj nitiatelection toirestorelonelECCS"RIR subsystes;ltheIcurrent; operational philosophy 3 s ;that;thi sf action;i s;j nitiatedtimmediatelyMThe CompletioniTimelofdamediately"ito: initiate:actionsithatlwould restorelatilea_st: one1ECCS;RlR;subsystenito10PERABLElstatus ensures;that1promptiactionfjsitaken;tolrestoreithe~ required coolingicapacitygRevi sing Jhe; CTSIActionatolimmediately initiate:; action ]j.s[ considered [an.;admini strati vef change Land ;j s consfstent;with_NUREGy1431G1 i CPSES Description of Changes to CTS 3M.S 6a 7/29/98

 -     _-           _ - _ - _  -            -                                                                    ]

W A L L s s s - s s s s s A e e e e e e e e Y Y Y Y Y Y Y Y Y C K E E R 5 Y T C 4

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I F L L s s s s s s s s 3 I O e e e e e e e e Y Y Y Y Y Y Y S B A W Y Y T C I t o T L P K A n H N P A E P s e 9 E d o n 0 R E H S E si o 3 e R C Si t Ph c S e U N A Cti r - C M O

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w-__----,,,---,--------------,.--------------e - - - ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.3-3 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 3-04 LG CTS 3.5.3 Action b ITS 3.5.3 Action B DOC 3-04 LG discussed two distinct changes. The first change involves a change in the wording of the Action requirement. The second change is movement of the instructions to maintain temperature using alternate heat removal methods to the Bases. Comment: The first change to the wording of the Action requirement should be separated out and justified as an "A" change. The movement of the instructions to the Bases is correctly justified as an "LG" change, but the justification provided in DOC [3-04 LG] is not adequate. Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG Response: DOC 3-04 LG has been separated into two DOCS ( DOC 3-04 LG and DOC 3-14-A.). DOC 3-04 LG now contains only the information related to the movement of information to the Bases. DOC 3-04 LG has been enhanced to provide the requested additionaljustification. New DOC 3-14-A was created to address the change to the wording in the Action requirement. ATTACHED PAGES: Enci 2 3/4 3.5-7 Enci 3A Sa and 6a Enci 3B 4 and Sa l l I l i< l i I.

 . . - - - - . - - .                                                                                                                                              I

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS T3,; < 350*F ECCS SUBSYSTEMS i LIMITED CONDITION FOR OPERATION 3.5.3.1 As : mini;;;, or,e One ECCS cubsystca c pri:cd of the fell; wing trajn ** 3 shall be OPERABLE:4 i usum n_ _ nnen.m,.c ms._,,..__, _u___ *MO

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d. An ^^E"""L" ' low path copeble ;f taking suction fre;;; the refu-ling water -tcr;;c tank upon bairg ;;,onaally re;ligacd er,d transfc ring

ucti;n to the contain.. cat :a;;;p during th recirculation pha:: of sp r;tien.

APPLICABILITY: H0DE 4. ACTION: i

a. WithnoECCScenttifugal[chtrgingpumpi(CCP)subsystemOPERABLE 3 13 LG j t__..__

uss..., v. 2_____ i6 ,t .~ . . ,vys i.. 2. ,. 2. . ,. . v, s . . .,u~ _, _ tu_ _ _ _ _ , , . . _ , mm us . n. . ... .. s..u...,,,,, 0 3.5.3 2 i pu;;;p er the flow path frer, the refuelitty W;tcr storage t nk, l resto.e at i #tt one ECf5 CCP subsystem to OPERABLE status within 3 03 LS 1 hour or be m COLD ShdlDOWN within the next-20 24 hours. 0 3.5.3 2 j

b. With no ECCS residual! heat removall(RlR) subsystem OPERABLE 3 04 LG becau: cf thi N;epei5ility of eithh t'hc rc-idual heat ic,,;;;1P 0 3.5.3 3 heat cxch;nscr er "l'", pu;;;p, immediately~;fnitiatelectioq3o restore .1 . . . .

at least one ECCS ltlR subsystem to OPERABLE status ee-meintain the 3 14.g Rc=ter C001=t Sy t= T,,, le:s th:n 350 F by u:c of altern:tc o.3.5.3-3 u_ . 4.s b . s.HV T . . __u_u f,N b, IV%5 J e A > - - - - -

• f "8 4 l A maximum of two charging pumps shall be OPERABLE capableicflinjectingiinto the_RCS whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F, except when Specification 3.4.8.3 is not applicable.

i f AFLRHR;trainjaayibe; considered 0PERABLElduring! alignment;andloperationi.for i decay;heatfamovalsificapablefofibeing;manuallyrealignedito;the1ECCS, mode Nb ofioperation? l

 - CPSESMark-up of CTS 3N.S                                                          3N S-7                                                                  7/29/98

CHANGE NUMBEP NSHC DESCRIPTION Times to reach MODE 5 from H0DE 4. Due.;to;the[ stable conditions;: associated;With; operation;Ein;. MODE;4Ethe 0 3.5.3 2 probability;of;occurrenceioffa1De'igniBasi_s Accident i sD owMAslairesul tTithe ;ECCS; operational. requirementsIare? reduced: with;onlyloneltraipTof;the;ECCS CCPLSubsysteeltequireditolbeJoperables2The1 required action 11 fit.he;CCESubsysteelj slLi .noperableli sItoj proceed , toicoldlshutdownj

                                                                            ~                                                           -

l 3 04 LG vensistent with ".U"EC lt31 Rev 1, che C:04 b ,\ l tarein;1;;y is revi cd. The requirc; cat to re-ter; at I least ena ECCS ubsysta; is revised to "i ;diately - lM initiets action to restore" an "J'", subsysta;;;. "ith both "J:",

pu;;;ps and Scot exchangcr3 inoperable, it wculd bc unwise to l requirc the plant to go to "00E S, whcrc the only available l

[ hcat rceeval systa;;; is the "s""s. Thercfore, the appropriate action is to initiets ;;osurc- to rcstcrc enc CCCS "Jl,"% sub y-tc; and to cor,tinu; the actions until the subsysta;;; is

.,__m___2 &_ mnen.og e _.g.._ , , _ _ mL. , ____&_ __ a __--_m

                                 , % d bus w%s bv vs E.rinut-b.         dbubude         nedv bauw u s bws g ru bb s wqw s u u,..uw s i k l,                                (if PPP, =nnct be restcred ) t =inteir T.,,4T--by use of   altcraats heat rc;;;; val ;cthods is d; criptive infor:stion

__; 2_ __.._2 .. .L. o ere ,L. .____J J__ ._ unne e J. usru ie virv v w%s bv burs vnwkw. sais bi ussd IL VIa kW 1 rvvh w se

                                 =1__.J..         ___LJLJ._J         J. A L .* _ ____._J.         L.. LL_     re a f*e usssuvy ysvigivebsv sss bli s e dwws uus su vy bits bvww spa-ification for "00ES 1, 2 and 3. CIS1315]3' Action _b                             .

Provideslan : alternate yequirementi(1 f;RHR2cannot; be restored)Ltoimaintaig1T,,, $350*F;by;use;off alternate 1her.t removalimethods ETheseideta11 slarelmoved;tolthelITS:Basesa Theseidet. ail s;.arelnotypecessary;tolens_ureithe1ECCSlare OPERABLE G The; requirements 1ofilTS;3f;5?3JLC0!and; Conditions are;adequateMotlensuringithat;thetECCSlare10PERABLEMThese deta1]siare;not1necessary.toiensure3helECCS can_.; perform their; intended;safetyifunction;M Asisuch Ethese deta11sia_re notfrequiredito[be;jn the3Sito;provideladequate; protection of;the; publ ic; heal.thiandisafetyRMoving; these; detail s maintains; consistency with;NUREGf1431 M Any;changeitojthese deta11sLw1111bejmadeljnllaccordancepith 10CFR50.59;.a.ndithe Bases; Control; Program; described 21n;ITSlSectioni515;14[ 1 I' CPSES Description of Changes to CTS 3M.S Sa 7/29/98 \ l

CHANGE NUMBER lE DESCRIPTION actuationlofcthegequired;ECCS1to,mitigatetthe 0 3.5.3 5 consequences loflatDBA; 3 11 LG Not applicable to CPKS. See Conversion Comparison Table (enclosure 3B). 3 12 A Not applicable to CPSES. See Conversion Comparison Table (enclosure 38). l 3113 LG CTSILCO 3:5j3; Action;aEterminology11s;rev1_ sed and.the 0 3'5'3 2 1 desctiptiyelinformation!j simoved ;toltheilTSlBasesl1 ) These[detailslare not;necessary1tolensureithe1ECCS] Required: Actions;are Leets;Jheirequirements;of;ITS;3;5131LC0 and : Conditons;areladequatelfor;ensurj ng ;the;ECCS; arej0PERAME7! Thesedeta1]sfareinotnecessaryltolensurelthe:ECCSican; perform 4 thei rfj ntended; safe.tyifunction &As t suc.h ttheseldetail s; ate not~ required;to be ggithe]TS)to! provide! adequate protectionjof the! publ ic1 heal thiandisafetyENovingithese" detail s maintains; consistency;with;NUREGi143.1 3 Anyl change tos these details 1wi1Rbe madelin;accordance:with 10CFR50;59 andithe , N Basest,ontroUProgram describedfin;ITS:'Section;5;5;L14; l

                                                                                                                                                                    /          3 s14                                                        A          CTSTLCOE3;5;3 Action bn providesn with'no ECCS:RlR --                                          1
                                                                       ~

subsystem 0PERABMtiheioptionitoleither[res. tore;at lO l least;one E_CCSIRHRIsubsystem2to10PERABLEistatusXorito maintainithe1RCSJT,4t 350Xby,u_se;of(alternate) heattremoval methodsEConditi.on ;Alof! ITS;LC013 i5j 3[requi resithatiwithino ECCS1RHRisubsystems10PERABLEithatiimmediate; action ; be injtiatedstoirestore aniECCSERHRlsubsystenitoiOPERABLE; status;3 Wh11elt.he; CTS [doesinott speci fyla;itietf;% tolinitiate; action to1 restore;oneiECCS;RlR;subsys t em;9%cyrenttoperational philosophy ;is;that;thi siaVjgj s11giti ated)1mmediately;EThe Comp.letionHime;ofilismediateMtolinitiate actions 1that*would restorelat3 east;one;ECCSilGRisubsystem:to;0PERABLEI status

                                                                                 -ensuresithatlprompt action!isitakenitoirestoreithelrequired cooljng;capacityURevisi.ngtthe;CTSTActionJto:immediately i niciate; action ii siconsideredfanj admini strative: change:and ii consistentfwithiNUREGi14315j                                                                      )

I l { 1 i CPSES Description of Changes to CTS 3N.S 6a 7/2988 l 1

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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.3-4 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 3-06 A CTS LCO 3.5.3 ITS LCO 3.5.3 Note This change is categorized as an administrative change even though it provides an exception to the LCO requirements that does not exist in the CTS. The DOC states that the note is only to

   " provide clarification."

Comment: Despite licensees'individualinterpretations of the CTS, the CTS themselves do not contain the allowance provided in the ITS Note. Therefore, this change should be reclassified as a less restrictive change and an appropriate justification provided. FLOG Response: As discussed during a telecon with NRC Staff on June 25,1998, the FLOG takes exception to this RAl. NRC accepted the same change at Vogtle as an administrative change, as discussed in Section 3.1.3.5 item (4) of the Vogtle SER wherein it was stated that this Note "is a necessary clarification when using the RHR system for cooling the RCS, when transitioning between MODES 4 and 5. Because this clarification constitutes existing operating practices, this change is administrative and is acceptable." in addition, the wording of CTS LCO 3.5.3.d, which refers to the RWST flow path "being manually realigned", supports the position that the new LCO Note, moved from the SR to the LCO per NRC-approved TSTF-90 Revision 1, is an administrative change . ATTACHED PAGES: None l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.3-5 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 3-10 LS-6 CTS 4.5.3.1.1 (DC), CTS 4.5.3.1 (All others) ITS 3.5.3 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: The NSHC for this change appears to provide the needed justification. Therefore, please incorporate the information contained in the NSHC into the subject DOC. FLOG Response: DOC 3-10 LS-6 has been revised to provide additionaljustification for the proposed change by adding the followiag information:

 "This change is acceptable because the ECCS operational requirements can be reduced due to the stable conditions associated with operation in MODE 4 and the decreased probability of occurrence of a Design Basis Accident (DBA). ECCS operational requirement reductions mean that certain automatic safety injection (SI) actuation signals are not available. However, in MODE 4, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA."

ATTACHED PAGES: Encl.3A 6 and 6a I 1 i l

                                                                                                                  )

L a

3 CHANGE NUMBER HSBC DESCRIPTION f 3 05 TR-2 Consistent with NUREG 1431 Rev 1, the requirement to submit a i Special Report within 90 days of an ECCS actuation and injection event is deleted. This change is acceptable because the requirement to submit a report is sufficiently addressed by the reporting requirements contained in 10CFR50.73. i 3 06 A Consistent with TSTF-90, a note is added to the LC0 that clarifies that an RHR subsystem's ECCS function is operable if l it is capable of being manually realigned to the ECCS mode of operation. This is an administrative change to provide { clarification. l i 3 07 H The surveillance frequency to verify a maximum of one charging l pump is capable of injecting into the RCS is changed from "at l least once per 31 days thereafter", to "at least once per 12 l hours thereafter". This change is more restrictive with regard to surveillance interval, however, the SR can be performed using control room indication and administrative controls. i 3 08 A A footnote is added to SR [4.5.3.1.1] indicating that the CTS SR to verify the RHR interlock action is not applicable when the RHR suction isolation valves are open to satisfy LC0 1 [3.4.8.3]. This LC0 permits the use of the RHR suction relief valves to satisfy the low temperature overpressure protection requirements in [ ] H0DE 4. When in this configuration, the RHR suction valves are required to be open. 3 09 H The surveillance frequency to verify a maximum of two charging pumps are capable of injecting into the RCS is changed from "within 4 hours after entering MODE 4 from H00E 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325"F, whichever occurs first and at least once per 31 days", to "once per 12 hours". This change is more restrictive l with regard to surveillance interval, however, the SR can be L performed using control room indication and administrative controls. 3 10 LS 6 Consistent with NUREG-1431 Rev 1 the requirement to demonstrate ECCS train operability in H0DE 4 in SR[4.5.3.1.1] has been revised to delete the 31 day surveillance to verify the correct position of each valve in the ECCS l DC ALL-001 flow path which is not already locked in place and the l' [18] month surveillance to verify automatic actuat, ion of ECCS nuanc =d automatic valvChisichangelis , ac ... lejbecausefthe;Ecczoperationantequirements

                                .anibe reduced duelto'theistable;conditionsiassocjated lo.3.5.35    g With;operationj in1HODEMianditheldecreased!probabil ity of(occuttenceloff aides.ign {Basj slAccidents(DBA)M ECCS operationalirequj rementj reductions;mean;that [certainlautomatic safety.3j njectionJ(SI)Iactuation; signal slarej potTavail able d             ;

However5jn;HODEM'Esuffjcientitimelexists fot manual CPSES Description of Changes to CTS 3N.S 6 7/29/98

CHANGE NUMBER RSliC DESCRIPTION 1 actuation 1ofithe;;requiredLECCSito.mitigateithe consequences;ofa!DBA? i - j 3 11 LG nu6 o g rc351e to CPSES. See Conversion Comparison iaoie (enclosure 38). 3 12 A Not applicabk to CPSES. See Conversion Comparison Table (enclosure 30). 3313 LG CTSiLC013 ; 5 :3 Metj onla l: termi nologylj sirevi sedlanRthe a 3's'3 2 i descriptiyelj nformation 51 s7 moved: tolthe;IIS; Bases;M These:detailsiareinotinec_essaryfi o; t ensure the:ECCS2 Required: Actions 1are~metMTheJfequirementsiof;ITS 3.5.3f LCO andlConditonslareladequat.elforlensuting1the2ECCSlare10PERABLEE Theseideta11 slareinot) necessary;to;;ensureithe.".ECCS~ can; perform theirlintendedisafety; function M Asisuch Rthese;deta11slare notirequited i.toibeij n ;therTSitoiproyj deladequ_ ate; protection;of theipubli.cihealthiand: safety;EZMoving,the.se detai_1s maintains 1 consistency;;with NUREGE1431 M Any;changeitoithese q details;will;beinadeJin'accordance,withj;10CFR50(59?andjthe l Bases 1Contro1Mrogram: described!1n;ITS1Section 5J5;14? 3?;14 A CTSiLCO 3;5J3; Action [h.);providesl1with;no;ECCStRHR . a-3.5.3 3 l subsy. sten 10PERABLEEthe; option 1toleitherirestote;at  ! least;one1ECCS!RHR; subsystem 1to;0PERABLEEstatusior;to maintain lthe;RCSjTg<^350*E byluse,ofial.ternateLheatl removal methopsl;1 Condition Atof2ITS LC013;5;31requiresithat:with;no . I ECCS; RHR;subsystensiOPERABLEi;that:jamedj ate; action! be l' jnitiateditolestorelag ECCS,RHR; subsystem;to 0PEPABLEistatus 3 WhjleithelCTSjdoesi notispecifylaitimelframeltotinitiatefl action totrestoreloneECCS1RlRsubsysten,1the1 current 1 operational Phil osophyj s j thatj thi st actionli sjj nitiated]j amediatel y;LThe CompletionlTjmelofi"jumediately'ito[initiateractionsithat would i restore:.atileastionelECCSlR % subsystenito 0?ERABLE; m status ensures;that:promptiaction11sitakenstoirestoreithe! required cool i ng;; capacity?S Rey 1 sing ;the1 CTS' Action 1to ;1mmedi ately i nit.1 ate;actj on? 1 s3onsideredf.anladmini strati ve; changelandli s consistent 1with NUREGF1431gf I 1 CPSES Description of Changes to CTS 3M.S 6a 7/29M8 L _ _..---._._--._-__.__._____-.D

I l I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.5.5-1 APPLICABILITY: CP, DC REQUEST: Section 3.4 DOC 6-21 LS-35 Section 3.5 JFD 3.5-4 CTS 3.4.5.2 Action b (CP) CTS 0.4.6.2 Action b (DC) ITS 3.5.5 Action A This change is a change to both the CTS and the STS and is beyond the scope of the conversion review and is generic. DOC 6-21 states that this change is consistent with WOG-84. Comment: Please provide the current status of WOG-84, if WOG-84 is not approved by the TSTF, then this change should be withdrawn from the conversion submittal at the time of the TSTF rejection. If WOG-84 has not been acted on by the TSTF, or is approved by the TSTF, but not approved by the NRC by the time the draft safety evaluation is being prepared, then it should be withdrawn from the conversion submittal at that time. This change will not be reviewed on a plant-specific basis. FLOG Response: DCPP and CPSES desire to continue to pursue the revisions proposed by this change. WOG-84 is now TSTF-236 which was approved by the TSTF on February 5, 1998. ATTACHED PAGES: , None 1

                                                                                                )

I i 1 I f i

l t ADDITIONAL INFO.RMATION COVER SHEET ADDITIONAL INFORMATION NO: CA-3.5-002 APPLICABILITY: CA, CP, DC, WC I REQUEST: Revise ITS 3.5.4 Bases to indicate that the RWST LCO, by virtue ofits temperature, l' volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident

analyses). I ATTACHED PAGES

Enci5B B 3.5-28 i

                                                                                                                                                                                                                                                          )

l l l l l l l l

B 3.5.4 BASES I t temperature. The upper temperature limit of E1991120*F is used in the small break LOCA analysis and containment ^^""A"!LI"I analysis. Exceeding this temperature will result in a higher peak clad "A"."I { temperature, because there is less heat transfer from the core to t the injected water for the small break LOCA and higher containment j pressures due to reduced containment spray cooling capacity. For j the containment response following an MSLB, the lower limit on { boron concentration and the upper limit on RWST water temperature l are used to maximize the total energy release to containment.

                                                                                ~

l The RWST satisfies Criter h 2:and)3 of 10CFR50;36(c)(2)(ii)] the l;;;, rc;1cy "tcts.st LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of i a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode. To be considered OPERABLE, the RWST must meet the water volume, I boron concentration, and temperature limits established in the SRs. APPLICABILITY- In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation, Core cooling requirements in MODE 5 are addressed by LC0 3.4.7, "RCS Loops-MODE 5 Loops Filled," and , LC0 3.4.8, "RCS Loops-MODE 5, Loops Not Filled." HODE 6 core l cooling requirements are addressed by LC0 3.9.5, " Residual Heat l Removal (RHR) and Coolant Circulation-High Water Level," and LC0 3.9.6, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level." l ACTIONS U i With RWST boron concentration or borated water temperature not within limits, they must be returned to within limits within 8 hours. Under these conditions neither the ECCS nor the Containment Spray System can perform its design function. CPSESMark-up ofNUREG-1431 Bases -ITS 3.5 B 3.5-28 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP-3.5-002 APPLICABILITY: CP REQUEST: CPSES has reversed its position of specifying minimum " indicated" % level for the RWST as stated in the CTS and instead is adopting the ISTS method of specifying minimum water volume in gallons. A Bases discussion has been added stating that an " indicated" level greater than 95% satisfies the surveillance requirement, but also allows the use of more accurate methods { for determining level. New DOC 05-5-A is used to address the CTS change. ATTACHED PAGES: Enci2 3/4 5-10  ; Enci 3A 8 ' Enci 3B 6 Enc!SA 3.5-11 Enci5B B 3.5-27 & 30 l i ( i l l ),

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITED CONDITION FOR OPERATION i 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A imum ' wt orated water kvolume;of 473,731; gall 95g 5 05 A CP 3.5 002

b. A boron concentration of between 2400 ppm and 2600 ppm of boron,

c. A minimum solution temperature of 40*F, and

d. A maximum solution temperature of 120*F.

APPLICABILITY: H0 DES 1, 2, 3, and 4. ACTION:

                                                                                                                                                                        )

Withithe RWSTiinoperableLdue to; boron.; concentration)or, borated;wateritemperature  ! "~" ' pot withinilimits,._ restore.;theitank;to;OPERABLEfstatus;within:8; hours 1or! b eiin at least.;HOTLSTANDBY, withinl th.einext 6 hoursiand;in' COLD SHUTDOWNiwithin the'fol_1owing 30; hours; With the RWST inoperable for; reasons;otherlthan; boron concentration oriboratedyater tse u temperature 1not withinjlimits} restore the tank to OPERABLE status within 1 hour or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS

                                                                                                                                                                         )

4.5.4 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) V r fying tie m4edieethoratedwater

2) Verifying the boron concentration of the water.

b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is less than 40 F or greater than 120 F.

l CPSESMark-up ofCTS 3N.S 3N S-10 7/29,98

l CHANGE NUMBER fiSliC DESCRIPTION restore boron concentration or borated water temperature to within limits is increased from 1 to 8 hours. Changes in boron concentration and temperature are slow and this revision provides a more reasonable time in which to restore limits. The contents of the tank are still available for injection during the completion time and the range of these parameters outside the limits is not likely to be significant to safety. 5-02 LS 10 The shutdown requirement for inoperable RWST is revised to require achieving MODE 3 within "the next" 6 hours. This effectively provides an additional hour to achieve MODE 3. The 7 hour total time to achieve MODE 3 is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. This change is consistent with NUREG-1431. Rev. 1. 5 03 A Not applicable to CPSES. See Conversion Comparison Table (enclosure 3B). 5 04 LS 12 Not applicable to CPSES. See Conversion Comparison Table (enclosure 3B). 5 05 A Con.sistent;with NUREG 1431;:the RWST; minimum;1evellis CP 3.5 002 now,expres sed ; i n; gall ons; i nstead ;of; percent! 1ndicated - JevelTc A^ discussion;off the1use ofdndjcatedilevelitu satisfylth_encorrespondingf1TS surve111ance3sjprovided;in:the Bases.J i 1 CPSES Description of Changes to CTS 3/4.5 8 7/29/98

W e SA - l r A A r u L - 2 - c L s s s 1 s s A e e e o - e e o of o Y Y Y N3 Y Y N N o N C o N t t r r e g a a p p n S K a yT yT S E h d d E c at at R e en en Y r e re 5 T C SA e . l r l r A r A r 4 I F - u u

   /   L    s              s            s 2                         - c                - c L                                                       1          s 3   I O  e              e            e                 o -               e             of      o          of     o B   Y              Y            Y                 N3                Y            N c      N         N o     N S   A W T   C I                   s ee T   L P

K A ot d o t o . nT t r N P A E P S n dA 0 a p E T s Ci i s d yT S R E H St h Sh i d at R C N E S e P v E t S P e en r e r U A C a h Cd u l A r u C M O s e oo t

                                         - l on c           s e

s e s e o

                                                                                                              - c of s

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     -   C Y              N n          Ni                Y                 Y            Y        N         N o     Y E

L t t t o t s B N O o n o n n o n i e t r A Y N s s s seoS . rm f ui o a S. p T A e oS dT e . oS dT e . oS dT dT h ot n r t yT C d N O P s P i P s P i P s Pi P s P i 8 o aS et i pT at en r e O L Ch Dt Ch Dt Ch D t Ch Dt h a yt T r d n l A r r S B

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                            - e v

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oa Nh oa Nh oa Nh Nh oa oe l N r ac u of N o Y e Y e N o A P r e s a r u n o i s m o e s M d1 a e 2k e h n s h o i t c 3 f r e r m o r e r o se u. O war o b no i i h a r u j e E D r e d i u q f t s nh ot C l n l on a o yc t P C . f o nnO i eM os a ne e r t n e r l l f t n C s p e aoS N sS iC n i u l f es na i h e t yom e er o or i c d l u m e r t o g no-nI O I dR o eet i i bl 2 a a1 l ob T fhf at t n ai r o w i u e m ii g sn hh a r v eg . s 4 t s q . i . d si S tt c eo4 pm k n ai 1 eiE n ni e T S el re t s r eud scn R i t of oe3 ms e h D cs W . s v nu sso at i r n t tdO nt R ee oo eip E h t ge r p mtT S oe t c o N . t r eM oi cmi e uo r ml u ih t rd p s V ni nv ua ped sl i Od a ug h cn nl l b h l k on l e8 xAr e N E G ot y i ca t ed I S yo h e v n ot i n C e Lt e t ei y s r e o r n oi a r6 e" T va t po mt o wE ol ; a O N A aj cn1 gcp a r t e .l c r 2 e i r n f et bh t pt o x St W n c1 neeB vh C H C i i 3 f if a nem idi d a r gc u 1. 4 n d rae e vsr ei r o w n e i n R e P r c N m O o i l I r set f r o y r ef n e r e t o r e P e Tf l df e ah o i r 3 e oet st oh C p C eesn C vel gt t e b tk f e f t' D A er vti E l s e 3 epS a a re t eat P eb a rt ce l p . T s t e r t en eu l cad h a c a n r s ou nn h e h t i s S t pi . eet mgdh eb r t t ei tl t a md a vd t t o on e m u a mh a r uno1 H s ceer r ooN ci vs et o . er et sv se mit v e N r t op rih y me ri ti ep i C rt Pi y pO oep-> r m r c i u r i p i w r u i m ntl r E i o u q C uc4 a t m u eq f e i nep N un t oL f j mp qr s e q3 v e i t mev; T O q eeyr .e it ni eins t e n d ce l Ti I rcie n rE oe or Trl i r r c met I a ns i6 o 0 oS h c or . D ch S e W p T b t Af t grt e i es nO t P e gs u w ea ad o e noin t t r wH e ea R' cnd r dh i t w ea u o go g w f t - I R C S n pq a men h ur o cpf i sS ri t i ae l p5 o p A s en v m ueet ol v hb or 4 p a mof a n t o i s mneog xsin e 'p tio l r l p md oe8 ct wo a o h d g t n ui h v s e nt a h s cn sl o n aT hS cW s R S oaa E Sdil P adi C ene E iC us ed e eceT ems ert eh i il i e es~ t1 D hC oo h oe h n e0 h uoo h o h c h a hh h nf TEh p TMb TibA T ppn Tb1 T a T g Tt Ti O R E B 0 2 M 3 0 4 0 5 0 6 1 7 21 3 41 5 _ U - - - 0 0 - S 0 - S 0 0 -

                                                                                                              - S 0

N 4H 4A 4M 4A 5L 5L 5A 5L 5A

l RWST 3.5.4 SURVEILLANCE REQUIREMENTS

                                                                 . =.:

SURVEILLANCE FREQUENCY

                                                                                                                           )

SR 3.5.4.1 .i f ;a- tj;d:--H NOTE 4-l[ -1 . lea'- $ 5 f 702 Only required to be performed when ambient

                                                                                                                   ;B-PS.:

air temperature is < 35;40fF or > 400 120 F.T3;

                      'afiEd:= 5      iit:'^a!?:1 5 % ";F;-"' = D i L .

Verify RWST borated water temperature is a 35 24 hours 40 F and s 400 120*F. SR 3.5.4.2 Verify RWST boratad water volume is a 455.200 7 days B PS gallon; M 73;731; gal 10 f CP 3.5 002 SR 3.5.4.3 Verify RWST boron concentration is 2 2000 7 days !B-PS/

                                                                                                                  ~

240_0 ppm and s E200 2600 ppm. CPSES Mark-up ofNUREG-1431 -ITS 3.5 3.5-11 7/29/98

l l RWST B 3.5.4 BASES APPLICABLE maximum boron concentration is an explicit assumption in the SAFETY ANALYSES inadvertent ECCS actuation analysis, although it is typically a (continued) non limiting event and the results are very insensitive to boron concentrations. Althoughlit;only has.La minor 1effect.1the maximum temperature is usedlinitheffeedline; break andismalllbreak;LOCA analyses;;casures that the amount of cooling provided frc; the RWST during the hcotup ph;sc of a fccdlinc brc d is consistent with safety :n lysis assu;ptions, the minimum temperature;is an assumption in both the HSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically non-limiting. l The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as 27 seconds, with offsite oower available, or 37 seconds without offsite power. This response time includes 2 seconds for electronics delay, a 15 second stroke time for the RWST valves, and a 10 second stroke time for the VCT valves. -Plants with a SIT nccd not be conccracd with the delay sincc th; SIT will supply highly bor;ted wcter pricr to RWST switchcVer provided the SIT is bctucca the pumps and the core. For a large break LOCA analysis, the minimum fontained ' h water volume limit of [455.200]G0,a; 473,731~ ons and CP 3.5 002 the lower boron concentration limn oT tee 003 2400 ppm are ""'""'" used to compute the post LOCA sump boron concentration necessary to assure suberiticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core. Thei_ limits;onL uinimum Teontained_ ' water volumelandinaximum boronJconcentration;of;the RWSTm also ensure 'aietttimum;maximumiequilibriunisumpl pH A 0.25)for CPb'5 0N thelsoluti_on; recirculated within1 containment;afterfa5LOCA which limits; corrosion:and; hydrogen productions The;11mit on maximum;boroniconcentrationjistalso useditoidetermine[aiminimum equilibrium sump:pH.1T_his; minimum pH Jeve);minimi.zes the evolution

ofd odineland minimizes the;effectiofichloridejstress corrosion'on i

mechanica1Jsystenslandl components? The upper limit on boron concentration of Ee2003 2600 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching f;om cold leg to hot leg injection is to avoid boron precipitation in the core following the accident. In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of E M } 40 F. If CPSES Mark-up ofNUREG-H31 Bases - ITS 3.5 B 3.5-27 7/29/98 l l

RWST B 3.5.4 BASES l The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST. With ambient air temperatures within the band, the RWST temperature should not exceed the limits. SR 3.5.4.2 The RWST water volume should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is normally stable and the contained,voluse requiredjis protected by an alarm, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experier.ce. Control;B. oar _d; indication . he CP 3.5 002 surveillance ~of'tha ra

                         %;cate<< 1 avol ' of)95%;;which qui rea.a includes nu i cated
                                                                            >5% !measur-+

RWST(waterivol ume h I uncertaintyAa conservative. verification;oficontained1volumeg

                              .er means;offsurveillance Dh.ich; consider; measurement 1 uncertainty maylal_so_be;u                     ~

SR 3.5.4.3 The boron concentratDn of the RWST should be verified every 7 days to be within the r' quired limits. This SR ensures that the reactor will remain subcritical following a LOCA. Fyrther, it assures that j the resulting sump pH will be ma'intained in an acceptable range so j that boron precipitation in the core will not occur and the effect of chlorlde and caustic stress corrosion on mechanical systems and components will be minimized. Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron concentration is appropriate and has been shown to be acceptable through operating experience. 1 REFER.ENCES 1. FSAR, Chapter E63 and Chapter E153. I l

       -                                                          #        ,.r-CPSES Mark-up ofNUREG-1431 Bases - ITS 3.5        B 3.5-30                               7/29/98 l

f 1 I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP-3.5-003 APPLICABILITY: CP REQUEST: The Bases description of SR 3.5.2.7 is clarified to included plant specific valves in the affected ECCS flow paths and their function. ATTACHED PAGES: Enci5B B 3.5-20 l l 1 i

ECCS -Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5 and SR 3.5.2.6 (continued) REQUIREMENTS Frequency is based on the need to perform tnese Surveillance under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillance were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program. SR 3.5.2.7 Re-The; correct; alignment of throttle valves in the ECCS flow path on an SI sional is necessary for proper ECCS performance, alves;.8810A,TB, C,10.are;provlaea in2 the charging pum cold leglinjection line_spValves 8822AEB,1 C,1Diare /p;tol ll CP.3.5 003 1 vided in the SI"onen to' cold" lea iniection lineg These manuallthrpttle valves arefpositioned following flow balancing and have mechanicalflocks steps to eHow ensure that the proper positioning for restricted flow to a ruptured cold legr casuring ,is, maintained (and;that the other cold legs receive at least the required minimum flow. This Surveill --- t not required for 9 5 unh ficw liniting crificesMlves,8816A?.8,,c,; D;a rovided .Li n the ;SI; pump, to .. hot . l eg Lrecricul ationjl i nes.; These manual throttle; valves?arel positioned following1 flow balancing ands have; mechanical locksitofensure; flow balancing'and;totlimit SI' pump runout! e 18 montn treguency is based on the same reasont as those stated in SR 3.5.2.5 and SR 3.5.2.6. SR 3.5.2.8 Periodic inspections of the containment sump suction inlet ensure , that it is unrestricted and stays in proper operating condition. l The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to have access to the location, and because of the potential for an unplanned transient if the Surveillance were l performed with the reactor at power. This Frequency has been found l to be sufficient to detect abnormal degradation and is confirmed by operating experience. (continued) CPSFRMark-up ofNUREG-1431 Bases -ITS 3.5 B 3.5-20 7/29/98

ADDITIONAL INFORMATION COVER SHEET , i ADDITIONAL INFORMATION NO: CP-3.5-004 APPLICABILITY: CP REQUEST: The Bases Applicable Safety Analyses description is revised to improve the discussion of the basis for the minimum and maximum equilibrium containment sump pH. ATTACHED FAGES: Enct5B B 3.5-27 l j l 1 i

l i RWST B 3.5.4 BASES i APPLICABLE maximum boron concentration is an explicit assumption in the ' SAFETY ANALYSES inadvertent ECCS actuation analysis, although it is typically a (continued) non limiting event and the results are very insensitive to boron concentrations. Althoughtitionly has;a minor 1effectl the maximum temperature 1.s;useds in:theLfeedlineL reak,and;small" b break:LOCA i analysesCca',ures that-the amount of eccling provided from the RUST l during the hcatup phasc of a feedline brcok is ccasistent with l safety analysis assumptions; the minimum temperature 21 s an assumption in both the HSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically non limiting. l The HSLB analysis has considered a delay associated with the ' interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as 27 seconds, with offsite . power available, or 37 seconds without offsite power. This response time includes 2 seconds for electronics delay, a 15 second j stroke time for the RWST valves, and a 10 second stroke time for i the VCT valves. I'lants with a SIT need not be concerned with the delay since the SIT will supply highly bcrated water prior to RWST switchovcr. provided the CIT is betucca the pumps and the core. For a large break LOCA analysis, the minimum contained water volume limit of [45C.200] 420,237 473,731; gallons and CP 3.5 002 the lower boron concentration limit of E20003 2400 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the i limiting case since the safety analysis assumes that all control l rods are out f core The l1mits o In1m ontained' water volume. an maximum ron concentration,o .RWST also

                                                                                                                     ]

q ensure a 6 n % m uax equilibrium l sump pH "" 0 N ti,e_ solution: recirculated within containment.2"for LOCA / CP 3.5 004 w ich limits corrosion _and hydrogen, production. ine 11mit on ' maximum ' boron . cone _entration ui s : al so used to idetermine ; a mi nimum equilibrium sump;pH. ThBNuinimusfptlevel minimizes tne evolution of 1odine:and'minimizesL thiieffect.of chloride; stress corrosion:on mechanicalEsystems;and components. The upper limit on boron concentration of E22003 2600 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident. In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of E%3 40*F. If CPSES Mark-up ofNUREG-1431 Bases -ITS 3.S B 3.S-27 7/29/98

ADDITIONAL INFORMATION COVER SrlEET ADDITIONAL INFORMATION NO: TR 3.5-001 APPLICABILITY: CA, CP, DC, WC REQUEST: Revise Traveler Status Gheet to; 1) reflect NRC approval of three travelers, TSTF-90, Rev.1, TSTF-117, and TSTF-153,2) delete reference to TSTF-155, which was rejected by TSTF and not incorporated by the FLOG, and 3) change WOG-84 to TSTF-236. There are no changes involved to any CTS mark-ups, ITS mark-ups, DOCS, or JFDs. ATTACHED PAGES: Encl.5A Traveler page I

                                                                                                                                                   )

l .__ _ _ - _______________________ w

1 l l l 1 I 1 l l l l INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.5 l l ( TRAVELER # STATUS DIFFERENCE # COMMENTS 1 l l TSTF 90, Rev 1 Incorporated 3.5 6 Approved lby;the NRCI m .3.5 001 l l TSTF 117 Incorporated 3.5 1 Approved lby the;NRCE TR 3.5 001 TSTF 153 Incorporated 3.5-8 epprovedibyJthe':NRC; m 3.5 001 TSTT-155 Net NA fict lRC appicycd as a 3.5 001 ir,ccrporated of traveler cut off date-W9G-64TSTFf236 Incorporated 3.5 4 DCPP and CPSES only. m 3.5 001 l l l l l l l

Attachment 4 to TXX-98182 Page 1 of 5 JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS CTS 3/4.9 -REFUELING OPERATIONS ITS 3.9 - REFUELING OPERATIONS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES w_-__-________-_________________. __ _ _ _ _ -

l l Attachment 4 to TXX-98182 Page 2 of 5 INDEX OF ADDITIONALINFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER 3.9.G-1 DC, CP, WC, CA YES 3.9-l DC, CP, WC, CA YES 3.9-1a DC, CP, WC, CA YES 3.9-1 b CA NA 3.9-2 CP YES 3.9-3 CP, WC, CA YES l 3.9-4 DC, CP, WC, CA YES 3.9-5 DC, CP, WC, CA YES 3.9-6 CA NA 3.9-7 DC, CP, WC, CA YES 3.9-8 DC, CP, WC, CA YES 3.9 DC,CP YES 3.'9-10 DC,CP YES 3.9-11 DC,CP YES 3.9-12 DC, CP, WC, CA YES 3.9-13 DC, CP, WC YES 3.9-14 DC, WC, CA NA 3.9-15 DC, WC, CA NA 3.9-16 DC, WC, CA NA 3.9-17 DC,CP YES 3.9-18 DC,CP YES 3.9-19 CA NA 3.9-20 WC NA 3.9-21 DC, CP, WC, CA YES 3.9-22 DC, CP, WC, CA YES 3.9-23 DC, CP, WC, CA YES 3.9-24 DC, CP, WC, CA YES 3.9-25 DC NA CP 3.9-001 CP YES CP 3.9-002 CP YES CP 3.9-003 CP YES CP 3.9-004 CP YES DC ALL-001 (3.9 changes only) DC NA DC ALL-003 (3.9 changes only) DC NA DC 3.9-ED DC NA TR 3.9-001 DC, CP, WC, CA YES TR 3.9-002 DC, CP, WC, CA YES TR 3.9-003 DC NA L

l Attachment 4 to TXX-98182 Page 3 of 5 INDEX OF ADDITIONAL INFORMATION (cont.) l ADDITIONAL INFORMATION APPLICABILITY ENCLOSEQ l' NUMBER WC 3.9-ED WC NA WC 3.9-001 WC NA l WC 3.9-002 WC NA WC 3.9-003 WC NA l WO 3.9-004 WC NA WC 3.9-006 CP, WC YES l l l i I-l I l l L_ _ _ _ _ _ _ _ _ _ _ _ _

Attachment 4 to TXX-98182 Page 4 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:

1. Each licensee is submitting a separate response for each section.

2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

3. If a licensee initiated change does not apply, "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,3B, l 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match  ;

the licensee specific conversion submittals. A licensee's FLOG response may not i address all applicable plants if there is insuffi: lent similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will ( be prefaced with a heading such as " PLANT SPECIFIC RESPONSE . ....." j

5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions, if the area being revised is not c! ear, the I affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.

6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.

7. . Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one or more of the other licensees. These changes are not included in the additional information for the licensee to which the I change does not apply, as the changes are only for consistency, do not technically l affect the request for that licensee, and are being provided in the additional information i being provided by the licensees for which the change is applicable. Tha emplete cet of changes for the license amendment request will be provided in a licensing amendment l request supplement to be provided later.

Attachrnent 4 to TXX-98182  ! Page 5 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued) I

8. The item numbers are formatted as follows: [ Source] [lTS Section]-[nnn]

Source = Q - NRC Question CA- AmerenUE DC-PG&E WC-WCNOC CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL"is used for the section number. I nnn = a three digit sequential number or ED (ED indicates editorial correction with no l impact on meaning) l I 1 I l i I e I i l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9.G-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.9.x Bases l General There have been a number of instances that the specific changes to the STS Bases are not properly identified with redline or strikeout marks. Comment: Perfom, an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal. FLOG response: The submitted ITS Bases markups for Section 3.9 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorialin nature and would not have affected the review. Examples of editorial changes are:

1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced.

2) Changing a verb from singular to plural by adding an "s" without " redlining" the "s".

3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g.,

SR3.6.6A.7).

4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference.

5) In some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.

6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.

7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was ,

determined to not be applicable, the text was then struck-out and remains in the ITS f Bases mark-up. i l l Differences of the above editorial nature will not be provided as attachments to this ' response. The pages requiring changes that are more than editorial and are not consistent with the markup methodology are attached. ATTACHED PAGES: None l l J

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O3.9-i APPLICABILITY: CA, CP, DC, WC REQUEST: General A great majority of the DOCS state that the reasons for the relocation and for the proposed changes including deletions, additions, and revisions, are made to be consistent with NUREG-1431. While this is a valid staternent, additionaljustifications are still required in order to support the proposed relocations and the CTS changes. The DOCS should be expanded to include additional justifications for the relocation and/or changes. Comment: Revise those DOCS which do not provide technical justifications for the proposed relocations and changes, and indicate which DOCS are being revised under this comment. FLOG Response: The statement that the changes are made to be consistent with NUREG-1431 in a number of the DOCS was not intended to be a justification for the proposed change but an indication that changes were being made to make the Technical Specifications similar to NUREG-1431. The DOCS were developed (specifically the Less Restrictive DOCS) with the intent that the No Significant Hazards Consideration would contain more detail justifying the change. The conversion license amendment application was developed using as a guide the Vogtle application for determining the level of detail needed for the DOCS. During the development of the conversion license amendment application in late 1996, several issues were identified with impact on the conversion process including " literal compliance," Generic Letter 96-01 and feedback from NRC concerning the acceptability of the submittals made by some other licensees. On January 24th,1997, senior managers from the FLOG met with the NRC (the Technical Specification Branch and Project Management) to discuss these and other issues. The utilities took additional time to review the conversion license amendment application to make sure that these issues were being properly addressed. On June 25,1998, a discussion was held with the NRC staff, in which it was believed that comments provided in Section 3.9 addressed those DOCS and JFDs that required additional justification. The FLOG has agreed to revise those DOCS. as identified by the NRC, by either bringing forward information from the No Significant Hazards Consideration (Enclosure 4) or l I providing additionaljustification as requested. No additional DOCS or JFDs have been revised unless indicated in the response to a specific Comment Number. I ATTACHED PAGES: None

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-1 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 3.9.1 DOC 1-01-A ITS 3.9.1, LCO 3.9.1 JFD 3.9-15

a. (Comanche Peak, Callaway, and Wolf Creek)

The CTS and ITS are proposed to be revised by adding "when connected" preceding

                     " Reactor Coolant System." The DOC provides a generic explanation, but it does not provide any specific technicaljustification for this addition. This revision is considered an administrative enhancement and a generic change to the ITS. Therefore, it must be reviewed and approved via the TSTF process before it may be adopted as the standard ITS language. Furthermore, Diablo Canyon does not include the proposed addition, "when connected,' in its CTS markup.

Comment: Either remove this item from the submittal and adopt the ITS language, or submit a TSTF for this generic change. Also, provide explanation why Diablo Canyon is not adopting the proposed language, "when connected." FLOG Response: The proposed changes to CTS 3.9.1 and ITS 3.9.1 were based on traveler WOG-103, Rev.1. WOG-103, Rev.1 has recently been designated TSTF-272 and transmitted to the NRC in May 1998. The proposed wording in TSTF-272 was modified from WOG-103, Rev.1, and these modifications have been incorporated into the ITS and ITS Bases. During preparation of the conversion license amendment application, WOG-103, Rev.1, was inadvertently omitted by Diablo Canyon. Diablo Canyon will incorporate TSTF-272 into the ITS and ITS Bases. ATTACHED PAGES: Encl. 5A Traveler Status page,3.9-1 Encl. 58 B 3.9-1, B 3.9-3, B 3.9-4 Encl.6A 3 l l L_-___-----____--___

Industry Travelers Applicable to Section 3.9 t TRAVELER # STATUS DIFFERENCE # COMMENTS l l l TSTF-20 Incorporated 3.9-2 Approved by NRC. TSTF-21; Rev 1 Not incorporated NA Appre;Td by 1R 3.9 001 NHC Change j not. consistent with purrent.p.lant operation. l TSTF-23, Rev. 93 Incorporated 3.9-13 Traveler bracketed ITS c 3,g 3 l 3.9.2 and revised the Bases for 3.9.3. l Bracketed Bases j j information from the { l traveler that is not j applicable to a specific plant ' was not incorporated. Approved byNRC. TSTF 51 Not incorporated NA Requires plant-specific reanalysis to establish decay time dependence for fuel l handling accident. TSTF-68, Rev. I Not incorporated NA Similar changes (Difference #3.9-1) were incorporated into the ITS based on current licensing basis. l l TSTF-92, Rev. I Not incorporated Not NRC approved as of traveler cut-off date. l l TSTF-96 Incorporated 3.9-4 Approved by NRC. l TSTF-136 Incorporated NA l TSTF-139 Incorporated NA TSTF-153 Not incorporated Net N!!C appmsd as of p1R 3.9 002 55mvbyt5 t us . v 5 5 date Editorial change.to 3.9.5 unnecessary whereas 3A and 3.5 were revised to match 3.9.5 wording WOG-76 Incorporated 3.9-11 mm

     ?.'00-103 ;-l )   Incorporated        3.9-15                                      ( 3,9,1, TSTF-272      /

m t ,

l l l Boron Concentration 3 9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LC0 3.9.1 Boron concentrations the Reactor Coolant System. san connected (rtions' of-? he refueling canal, and the refu ng 3.9 15 cavity shall o maintained within the limit specified in the 0 3.9-la COLR. _ ti'!!'? ?iE*iiiidiEE ?1iNOTE!Bt 2 5-l -? gh!;ufi m i:yyy y y While this;LC0;is;not' met;ientrylintofMODE 6;from H0DE1511s not g3*g,y4q permitted;' . 7 g m. ; ;._.,.c.,u.  ;;;;;;. .;;.;,=,;;.t;;  ; s; ;; .,  ;;;;;;;; s,;;;; ;; ;; .,,. .; .;;w ;,,;;.;;;  ;;.;;.; e u 4 ..s., 2 APPLICABILITY: MODE 6. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE ALTERATIONS. Immediately within limit. MiQ A.2 Suspend positive reactivity additions. Immediately MiQ A.3 Initiate action to restore boron concentration to Immediately within limit. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l l SR 3.9.1.1 Verify boron concentration is within the limit 72 hours ,

ED~ ,

specified in the COLR. ' CPSESMarkup ofNUREG-1431 -ITS 3.9 3.9-1 72 9/98 L

Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the Reactor Coolant System (RCS), the connected.portionstof3erueling canal, and 0 3.9-la the refueling cavity "Ing refueling ensures that the reactor remains subcritical during MODE 6. Refueling boron concentration is the soluble boron concentration in the coolant in each of these volumes having direct access to the reactor core during refueling. The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling boron ) concentration limit is specified in the COLR. Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of k,,, s 0.95 during fuel handling, with ' control rods and fuel assemblies assumed to be in the most I adverse configuration (least negative reactivity) allowed by plant procedures. GDC 26 of 10 CFR 50, Appendix A requires that two independent reactivity control systems of different design principles be provided (Ref. 1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) is the ina.1_nisystem capable of maintaining the reactor suberitical in cold conditions > by maintaining the boron concentration. i The reactor is brought to shutdown conditions before beginning i operations to open the reactor vessel for refueling. After the  ! RCS is cooled and depressurized and the vessel head is unbolted, i the head is slowly removed to form the refueling cavity. The refueling canal and the refueling cavity are then flooded with j borated water from the refueling water storage tank through the  ! ! open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps.  ! The pumping action of the RHR System in the RCS and the natural j circulation due to thermal driving heads in the reactor vessel (continued) ?  ! CPSESMarkup ofNUREG-1431 Bases -ITS 3.9 B 3.9-1 7/29/98

                                                                                                                              ---._-_-.-_________-____J

I . I l I l Boron Concentration B 3.9.1 1 I BASES , i LC0 The LC0 requires that a minimum uniform-boron concentration CP 3.9 001 be maintained in the filled; portions;ofdthe;RCS whjle in H0DE"6: " Additionally,5.whentthe RCS' is c flooded'andQ 0 3&la

                                              @nnecteEto the refueling danal, andIhe refueling caviD Ine same minimum; boron; concentration;1strequ_1 red,tojbe maintainediin theitilledl portions offthegefueling: canal;and refuelin;/;cavityphile ir. "00E 5. The boron concentration limit specified in the COLR ensures that a core k,y of s 0.95 is maintained during fuel handling operations.

l Violation of the LCO could lead to an inadvertent criticality during MODE 6. I APPLICABILITY This LC0 is applicable in H0DE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a k en s 0.95. Above MODE 6. LC0 3.1.1,

                                               " SHUTDOWN MARGIN (SDM)-Tm        .. ,   ...m.   ....m.    . . . . . - . .

MARGIN (SD") Tm 200"f " LC013.1_.5;1" Shutdown Bank Insertion Limits,"Jand;LC0 3;1,671 Control;Bankunsertion Limits Eensure that an adequate amount of negative reactivity is available to shut down the reactor and maintain it subcritical. Thelapplicab111ty_is modified by a:Noteistating;that1 transition fr.om H00E;5;to,HODEL611sinot, permitted. ZThisl Note specifies;an exceptjon'to LC0 3;0:4Jand prohibitsithe. transition when boro _n concentrationJimitsareinotmet,EThis; note;assur.es;that.; core reactivity 21s_majntainedLwithjnslimits;duringifuelihandling operations] ACTION 3 A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LC0. If the boron concentration nf any coolant volume in the RCS, and 0 3.9-la connected ions'.o the refueling canal or the l refueling cavi@y. t is iess than its limit, all operations I involving CORE ALTERATIONS or positive reactivity additions must ( be suspended immediately (continued) CPSES Markup ofNUREG-1431 Bases -ITS 3.9 B 3.9-3 7/29/98

Boron Concentration B 3.9.1 l BASES Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. When; determining ;compl i.ance:with ; actions;;; addition;of; borated wateriwith;alconcentration; greater:thanLor;equalito the, minimum r.equired RWST; con. centration;shal]Lnot;be:consi.deredia positive reactivity;changeG(RefZ3) bl In addition to immediately suspending CORE ALTERATIONS or positive reactivity additions, boration to restore the concentration must be initiated immediately. In determining the required combination of boration flow rate and l. concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration. i SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentr inithe fjlled portjonsfofithel.RCS, i an connect 'rtio l (1(e;,:having; direct access to t.e:reactorivesse ,, ell;the 0 3.9 la- I fnlei g G 3 R of;the refueling canal and the refueling I cavity, tare-+s within the COLR limits. The boron concentration of the coolant in each volume is determined periodically by chemical analysis. A minimum Frequency of once every 72 hours is a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based on operating experience, which has shown 72 hours to be adequate. l CPSESMarkup ofNUREG-1431 Bases -ITS3.9 B 3.9-4 7/29/98 L_____---__.______

CHANGE NUMBER JUSTIFICATION 3.9 13 Not applicable to CPSES. See Conversion Comparison Table (Enclosure 6B). 3.9 14 A MODE change restriction is added to ITS 3.9.1 in the LC0 Applicability per the matrix discussed in CN 102-LS 1 of the 3.0 package. This restriction prevents a transition from MODE 5 to MODE 6 if the boron concentration limit for MODE 6 is not met. (See LS 1 NSHC in the CTS Section 3/4.0, ITS Section 3.0 package). LC013;.0.4'hasi.been revisedisofthatichanges?in N0 DES or.;other;specified; conditions';in the: Applicability 1thatlare 0 3.9-24 part offa: shutdown;of;theiunit;shallinot be prevented. ;In = additionKLC013.0i4 has;been revised soit. i hat ~1t;.1s;only appli cabl e;forlentrylintof a' H0 del orf other ' speci fied _ conditions 2in { thel.ApplicabilityminMODES11J2,;3,:andA.LThe1 MODE 1 change j restrictions?inJLC0 3.0;Owere previouslyiappljcablefin all i MODES? LITS:LC013;9.1;was;modifipd byia1 Note, stating:5"While;this LC0 ~ 1 s not met tentryf.i nto ; MODE % from , MODE L S.ii s L not l permi_ tted . "2 Required. Actions woul_d. prevent the Defueled;;ModeL6ftransition j byisuspending; CORE ALTERATIONSfand.'positiveLreactivityiadditions.L ' The;transitionffrom_Modei5ito. Mode 161could. occur without adequate , boration!for; MODE; 6 fequi rements .

                                                                                                                                          ]

3.9 15 0-3.9 la l o ? 01 hn . een revised in accordance

                                        -103 TSTF 272 o clarify that boron concentration limitsTo with    travelerh not apply i.o the refueling [ cavity and refueling canal] or other flooded areas when these areas are not connected to the RCS.

This change is acceptable because the boron concentration limit is intended to ensure that the reactor remains subcritical in MODE 6. However, when areas containing boron solution are isolated from the RCS, no potential for boron dilution exists. Therefore, there is no need to place a limit on boron concentration in these areas when they are not connected to the l RCS. This change is consistent with the intent of the Specification, as described in the Bases, and eliminates I restrictions titat have no effect on safety. i CPSES Differencesfrom NUREG-1431 -ITS 3.9 3 -7f29/98 i i

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-2 APPLICABILITY: CP REQUEST: CTS 3.9.1 Action b DOC 1-11-LS19 ITS 3.9.2 Action A4 The time required to verify that the boron concentration is within its limits has been changed from 1 hour to 4 hours in CTS. While "4 hours" is consistent with ITS, the explanation that was provided, "The 4 hour requirement is a reasonable estimate of the time requirement to measure the boron concentration by chemical analysis," does not address any technical justification for this change. Comment: Revise DOC by providing specific technical justification for this relaxation and the impact which may affect CPSES in terms of plant

• operation, design and licensing basis.

FLOG Response: DOC 1-11-LS-19 has been revised to provide specific technical - justification for this relaxation and the impact which may affect CPSES in terms of plant operation, design and licensing basis. ATTACHED PAGES: Enci 3A 2 & 2a l i i l I L___--------------_---___-----------.------------__----_---.-- - - - - - - - - - - - - --- - - - - - - - - - 2

CHANGE NUMBER RSliC DESCRIPTION 1 06 LS-1 The requirements to initiate boration at a specified flow rate having a specified boron concentration is replaced by the more general requirement to initiate boration to restore the required boron concentration. The reactor operators are expected to select the best method of increasing the boron concentration to the required value, specified in the COLR. The proposed change clarifies that action is applicable only to restoring boron concentration to within limit. This change is acceptable because it is an example of removing procedural details while maintaining the actual limiting condition as a TS requirement. 1 07 H Nct applicable to CPSES. See conversion comparison table

                                    -     (enclosure 38),

1-08 H Separate entry into the action is allowed for each unborated water source isolation valve. The only requirement affected by this note is the requirement to verify reactivity condition. A note is also added that boron concentration verification must be completed whenever this action is entered. The overall impact is more frequent verification of reactivity conditions. This change adds a more stringent TS requirement which is appropriate and consistent with NUREG 1431, Rev. 1. 1 09 LS 2 The SR to verify reactivity conditions is deleted, as it is generally descriptive of the Mode 6 conditions, as defined in NUREG 1431, Rev 1, and is addressed by SR 4.9.1.2. This change is acceptable because the boron concentration is required to be within limit prior to entry into MODE 6 in accordance with the Applicability Note for ITS 3.9.1. Thus, the deleted SR is redundant to other requirements that remain in TS. 1 10 LG Hoves the description in the SR to determine the boron concentration by chemical analysis to the BASES. This change is consistent with NUREG 1431. Rev. 1, and removes details that are not required to be in the TS to protect the health and safety of the public while retaining the basic limiting conditions for operation. 1 11 LS 19 Consistent with NUREG 1431, Rev. 1, the time required to ! verify that the boron concentration is within its limits has been relaxed from I hour to 4 hours. The 4 hour requirement is a reasonable estimate of the time requirement to measure the boron concentration by chemical analysis. The; likelihood of;a;significant o.3.9 2 reduction;inithe boron concentration during. MODE;6 CPSES Description of Changes to CTS 3N.9 2 7/29/98 u--- _ _ ___ ________ _____ ___ _

CHANGE NUMBER RSliC DESCRIPTION operationssis;smaltdue;.to,theLlargell mass;of borated 0 3.9 2 wateGin . the..;refuelj ng l cavity;andithe; fact ;tha.tiall unborated Water; sources 1aredsol_atedl:precludingJ dilutionLHhe boronJconcegration;is; checked;every 72thours;durj ng 30DE: 6;undet;SR ,3.;9,1712Hhelrelaxation to;41 hours proyjdes:aimoretrea.sonableitime:tolobtain and analyze"an"RCS: sample;for boroniconcentration;and_basedion thelimmediate; action; to1close; the [affected : val.yes ;Twoul d not belcause]aLfurther;; decrease;in boron;concentrationtif al dilution event 1hadloccurredi 1 12 LG Generalizes the requirement to verify the dilution isolation valves are closed by mechanical stops or removal j l, i 1 l 1 l l l 1 l CPSES Description of Changes to CTS 3/4.9 la 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-3 APPLICABILITY: CA, CP, WC REQUEST: CTS 3.9.2 DOC 2-01-LS DOC 2-01-LS-21 (Wolf Creek) The requirements related to indication provided by the source range detectors,"each with continuous visual indication the control room and one with audible indication in the containment and control room", is proposed to be deleted from the LCO of CTS in accordance to NUREG-1431, Rev 1 and TSTF-23. While TSTF-23 is still under review, it cannot be adopted until it is approved. Comment: Revise the submittal by including the above phrase or provide further technical justification to support the proposed deletion. FLOG Response: The latest status report from the TSTF industry database, dated June 16, 1998, indicates that the NRC has approved TSTF-23, Rev. 3. The FLOG continues to pursue the changes approved in TSTF-23, Rev. 3. Additional information supporting this DOC is in NSHC LS-21 in Enclosure 4. DOC 2-01-LS-21 is revised to include the following information:

                                                                  "In Mode 6, the source range monitors are required for indication only and there are no precise setpoints associated with these instruments. In this capacity, the source range instrumentation is typically used to read a relative change in count rate. The source range instrumentation is rnonitored for significant changes in count rate which are impor1 ant to evaluate the change in core status. The accepted convention for defining criticality does not require precise or specific setpoints or indication, but only requires verification of a slowly increasing count rate.

Consistent with NUREG-1431, Rev.1, the Technical Specification requirements consist of maintaining two source range neutron flux monitors OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. There is no requirement for an audible signal or alarm to initiate operator response oecausu in MODE 6 reactivity changes would be slow and a boron dilution accident is not postulated. The occurrence of a boron dilution event is precluded by maintaining the isolation valves from unborated water sources secured in the closed position per ITS 3.9.2." ATTACHED PAGES: Encl.3A 3 Encl. 5A Traveler Status page l l

CHANGE NUMBER RSBC DESCRIPTION

of motive power. Consistent with NUREG 1431, Rev. 1, the valves are only required to be " secured." The specific method of securing each valve will be contained in plant l procedures.

2 01 LS 21 Consistent with NUREG 1431, Rev. 1 and TSTF 23. the requirements related to indication provided by the source range detectors would be deleted from the LCO. In accordance with TSTF 23, the requirements for visual indication for plants that do not rely on a boron dilution analysis would be discussed in the Bases; while the requirements for audible indication would be eliminated as a Technical Specification requirement. o.3.9 3 In; ModeJ6:,'Ithelsource;tangelsonitors *are;requj redifor jndication;oplylandithere:are:nolpreciselsetpoints associated $1th these;instrumentsMInitt!1sicapacityEthe sourceirangeMustrumentetionlj sl:typicallfuseditoltead;a relatlye;changeZ1M countirate 2 Thelsource fange j instrumentation;i slaonitorediforlsigni ficantichangeslin countiratelwhich areLimpor, tant;to; evaluate 1the[changelin coteistatus m The"acceptediconventioniforldefining cr.iticalityldoeRnot: requj re! Preci se;or: spectfj cisetpoints oGndicationEbutionly;tequites;verificationlof;a;siowly increasing;countirate & Consistentiwith NUREGi1431% Rev? IEtheJechnicallSpecif,1 cation; requirements consist;of maintaining;twolsourceirangelneutrontflux;eonitors OPERABLEltolensureithatIredundantimonitoring;capabilityiis ava11ableEtoidetectichangesiin coreTeactivityRThere31s

polrequirementfonLanlaudibleisignallorlalare;tolinitiate I operator

response 1because;1pl M00E;61reacti vity; changes woul_djbels1owlandla;boronidilutjon7accidentjisJnot postul ated aTheloccurrencelofca? boronidil utionteventiis -

precl udediby1 maintaining 1theli sol ationival vesifrom unborated water 2 sources

• secured indthelclosedl position; pet ITSf3;9;2. This change is acceptable because it would l eliminate requirements associated with indication channels that are not required to mitigate boron dilution events.

l 2 02 H The ACTION statement is revised to require that l restoration of one monitor is immediately initiated. This change adds a more stringent TS requirement which is

appropriate and consistent with NUREG 1431. Rev. 1.

2 03 LS 3 The ACOT requirements are deleted and a Channel Calibration is added, in accordance with NUREG 1431, Rev.

1. In Mode 6 the source range monitors are required for indication only and there are no precise setpoints ,

associated with these instruments. In this capacity, the  ! source range instrumentation is typically used to read a i relative change in count rate. The source range instrumentation is monitored for significant changes in i CPSES Description of Changes to CTS 3N.9 3 7/29/98

l Industry Travelers Applicable to Section 3.9 l 1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-20 Incorporated 3.9-2 Approved by NRC. TSTF-21; Rev 1 Not incorporated NA Approved by 1R 3.9 001 NRCChange imnummmmum niet consistent with current plant operation.

                                                                                                             ,             l TSTF-23, Rev
                     ~

Incorporated 3.9-13 Traveler bracketed ITS ( 3.9 3 3.9.2 and revised the qumammmmmum Bases for 3.9.3. Bracket,d Bases I information from the traveler that is not applicable to a specific pla t was =a* t--arated. g kreved by"NR TSTF-51 Not incorporated NA hqunes plaIpecific reanalysis to establish decay time dependence for fuel handling accident. TSTF-68, Rev. I Not incorporated NA Similar changes

 '                                                              (Difference #3.9-1) were incorporated into the ITS based on current licensing basis.

! TSTF-92, Rev.1 Not incorporated Not NRC approved as of traveler cut-off date. TSTF-96 Incorporated 3.9-4 Approved by NRC. TSTF-136 Incorporated NA l TSTF-139 Incorporated NA ! TSTF-153 Not incorporated Net N"C approved ;; ef ag.3.9 002 trav;;;r ;;; s'f date. Editorial change to 3.9.5 unnecessary whereas 3.4 a.ad;3.5 were revised to matchf3.9.5; wording WOG-76 Incorporated 3.9-11

      "/OC-103g             Incorporated         3.9-15                                                         ( 3.9 la TSTF-272

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-4 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 4.9.2 b and c CTS 4.9.2 b, c and Footnote * (Diablo Canyon) DOC 2-03-LS3 Surveillance requirements 4.9.2 b and c for Analog Channel Operational Test are proposed to be deleted in CTS to be consistent with NUREG-1431, Rev.1. ITS does not include these requirements. DOC 2-03-LS3 discusses the reasons for deletion, but it does not address the associated impact in regard to plant operation and design basis, and whether these surveillance would be moved to plant procedures or relocated to the UFSAR. Comment: Revise DOC to justify as to why this is acceptable based on licensing and design basis. If these SRs should be relocated, identify the plant document that includes the CTS requirements. FLOG Response: Additional information supporting this DOC is in NSHC LS-3 in Enclosure

4. As discussed with the NRC technical specification branch reviewers on June 25,1998, DOC 2-03-LS-3 is revised to include the following information:

                                                    " During REFUELING, the source range monitors provide visual () indication of neutron count rate to plant operators. Core reactivity is maintained primarily by the requirements of ITS 3.9.1

[and 3.9.2] which assure that the boron concentration in refueling water is within limit and that dilution of the boron will[not occur]. Thus the neutron monitoring channels provide further assurance that criticality will not occur. The proposed deletion of ACOTs for these channels would be offset by the CHANNEL CHECK and CHANNEL CALIBRATION requirements. The addition of a CHANNEL CALIBRATION to be performed every 18 months provides assurance that the instruments can provide the visual indication. There are no alarms, interlocks, or trip setpoints associated with these channels that are required to be OPERABLE during MODE 6. In addition, in MODE 6 the source range instruments provide no automatic actuation function used for mitigation of accidents, and they would have no effect on the outccme of an accident. Furthermore, the modification of SRs for these indicators does not imply that they will be unavailable when required. The CHANNEL CHECKS and CHANNEL CALIBRATION SR that remain in effect provide the necessary assurance of OPERABILITY. " ATTACHED PAGES: Encl.3A 3a I l l L-__ - - _ - - . - - _ . - - - - _ - - - - - - - - - - -

CHANGE l NUMBER N2iG DESCRIPTION i count rate which are important to evaluate the change in i core status. Even the accepted convention defining criticality only requires a slowly increasing count rate be verified. Consistent with NUREG 1431, Rev. 1 indicating instruments only require channel checks and channel calibrations. The more frequent ACOTs are applied only to those channels with operational interlocks or other setpoint actuations. Therefore , the MODE 6 channel checks and channel calibration requirements for the source range monitors are adequate to assure their operability, considering the more frequent ACOTs performed on this instrumentation in other Modes, the effectiveness of these surveillance requirements in maintaining other indicating instruments operable, and the accuracy required of theseinstrumentsinMode6.During3EFUELINGCthe sourceirangeimonitorsJ provide; visuali[3 ; indication : of 0 3.9 4 peutron: count rateito pl.antioperatorsG; Core reactiv.ity isimaintained;ptimarily;by;the; requirements.;otITS 3.9;1?[and 3,9.2]iwhichl:assurelthat;the; boron concentration;in refueling water?is;withinflinjtTand;that dilutioniofxtheiboronlW111l[notfoccur] C Thusitheineutron monitori ng_ channel si provideifurtherfassurance;that criticality willinotloccurDThe;proposedideletioniof ACOTsiforlthese:channelsiwould;beloffset;bytthe; CHANNEL CHECKlandiCHANNEL;CALIBRATIONLrequirementsGThe addition offa CHANNELECALI8 RATION to;belperformedievery:18foonths providesIassuranceithat<,t_heinstruments;canlproyfde;the visushindication.;EThereareinoialarmssinterlocks&or tripisetpointsiassociated;withthese; L channels;thatTare requiredito;be;0PERABLEIduring1HODE16 M !niaddi. tion ein MODEi6;thelsourceirange;instrumentsprovideinolautomatic actuationjfunctioniuse(for;nitigationjoffacciden.tstand theyi;wouldihaveinoieffaction theloutcomeiofian; accident;; m j Furthermore 6thelmodi ficationioGSR.siforithese;ind1cators l doesinotligly thatithev >yillibelunavailablelwhen I required M The: CHANNEL % nECKSlandlCHANNELICALIBRATION"SR l that remain;1n;effect; provide;the necessary; assurance:of OPERABILITY. 2 04 LG Not applicable to CPSES. See conversion comparison table (enclosure 38). 3 01 R Consistent with NUREG-1431. Rev. 1, the suberiticality requirement prior to irradiated fuel movement is relocated l 6 i i CPSES Description of Changes to CTS 3N.9 3a 7/29/98 L-_--_ _ _ _ - - _ - - _ - - - - - - - - - - - - - _ - - - - - - - -

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-5 APPLICABILITY: CA, CP, DC, WC l REQUEST: CTS 3/4.9.3 l DOC 3-01-R The CTS requirements in 3/4.9.3 have been entirely relocated to an unspecified licensee controlled document. Though Conversion Comparison Table provides the new location of this l item, it is still necessary to address where the CTS requirements have been relocated to in the ! DOC. In addition, the specific technical justification fer the relocation is not addressed in the DOC.

Comment

Revise DOC by providing justification as to why the relocation is acceptable and identify the licensee controlled document to which the CTS requirements would be relocated.

FLOG Response: DOC 3-01-R has been revised to provide justification as to why the relocation is acceptable. The justification shows that this LCO ensures the decay of short lived fission products prior handling irradiated fuel. The associated decay time is not an installed instrument nor is it used to detect an abnormal degraded condition. This decay time is an initial condition of a DBA which assumes the failure of fission product barrier. The industry /NRC did, however, agree during development of NUREG-1431 that this LCO could be relocated and only included it in the Bases of NUREG-1431 (B 3.9.7). The decay time is not a structure, system, or component of the primary success path to mitigate a DBA. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998. ATTACHED PAGES: Enci 3A 4 Att 21 39 & 40 1

CHANGE NUMBER NSBC DESCRIPTION

f. to a licensee controlled document. JhjsiCTS requirement 0 3.9 5 l ensuredithatisufficientitime;had[ elapsed;to;allowjthe l- radioactive 1 decay;offshortil j vodifj ssion:productsJ priot l toleovementiofiittadiatedifuelgiThis change is acceptable based on the schedule requirements following shutdown to attain plant conditions for movement of irradiated fuel.

l These schedule requirements provide assurance that the requirements of the decay time LC0 would not be exceeded. Thi s; proposediTS;revj sjon; relocates; requirements EwhjcKdo l potimeetitheiTS;ctitatiallpl10CFR50[36(c)(2).(11)Eto documents 1withles.tabl ishedicontroll prograns OThis

regulatjon: addresses 1the{scopond; purpose;ofgSmIn l .dojng so1T;1t sets;fotthtalspecificiset!oflobjectjve l ctiterf a ;fotideterminingMichlregul atory; requirements :and )

operatingiresttjctionsishould:belipcludedlinithelTSI . Relocation 3fithese; requirements!allowsithelTSitolbe L reservedionly!forithosaiconditions;otI]imi.tationslupon teactorloperation;whichlare;necessaryitolobviate;the possibil itylotan[abnormellsituationlorleventig1ying: rtse to!anlimmedjate;threatitolthelpublic;healthiandisafety thereby1 focusing 1thescopelofLtheTSHAnievaluationlofithe applj cabilj tyZof;these;crjtetjaitoithis t specifjc3 tion (1s l provided5jn' Attachments 21] j ! l ' l Io; ensure; aniapproptjate11 eve 1Lof;controM;these requitementsy1]Eb(telocatedito:1)1 documents 1that;are subjectitolthe[ptoVisjonslofil0iCFR;50j 59 E2);other licenseeldocumentslwhichihaveisimilatiregulatorEcontrol s (e;gTfithe;Que]Jty: Assurance 21an,Tasidescribeds;in;the ESAREWhich;isicontro11edi by110CFR50;54a)Ror13)(to programs;thatrare:contro11 M yiaithe; Administrative l Controlslsectionioffthe:Japroved:ISnThe;1 identification i ofithelspecifiellicenseelcontrolled[ document;containing thisitequirementiis!Provided;1n; Enclosure 13B;ofithe - confersion;submittall] Compliance:withithelrelocatedirequirementsiwil1[not;be J

                                   'ffectedbyithis; proposed [changeltolthe[ current 1 Technical a          3 Specification.saThelrequited petiodicisurvetljances:will continueltoibejperformedito; ensure;that:11mitslon parameters:ar.e  m maintained 3 Iherefore trelocationlofdthese         )

i l requirements 1wi1Ehave nojimpactiontsystenloperabilitylot the: maintenance;of; controlled;paraseters;withinTlimits; I 4 01 LG This change removes the word " automatic" from the requirement that each penetration be capable of being closed by an OPERABLE automatic containment [ ventilation] isolation valve. The requirement for the automatic valve would be stated in the Bases. This change is consistent with NUREG 1431, Rev.1. and removes details th.at are not i i crses nescripaon of changes so crs 3/t.s t 1/2siss

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.9.3 DECAY TIME Applicable MODES: During movement of irradiated fuel in the reactor vessel (2) EVALUATION OF POLICY STATEMENT CRITERIA ls ths Technical Specification applicable to: YES NO _, 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. i _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION I This specification places a time limit on reactor subcriticality prior to the movement of irradiated fuel assemblies in the reactor vessel. This ensures that sufficient time has elapsed for the radioactive decay of short-lived fission products. The decay of short-lived fission products is assumed in the fuel handling accident. Decay time is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. Decay time does not satisfy criterion 1. Decay time is an in:tial condition of a DBA that assumes the failure of the integrity of a fission product barrier. However, it was agreed t;pon in Industry /NRC meetings during the development of NUREG-1431 that this LCO may be relocated. The assumed basis for j not retaining the accident analysis assumption is that it is not "an operating restriction." It l Is not a necessary constraint because plant design and other limitations are more restrictive. it is clear that not all accident analysis assumptions are required to be included in TS [TS primarily exclude physical design features); therefore, the basis for exclusion is that the operator does not need to control the parameter. This LCO is not contained in Ref. 2. However, the requirement for a minimum decay time of 100 hours prior to fuel handling is contained in the Bases of NUREG-1431, Rev.1 (B 3.9.7). CPSES will be consistent with the decay time limit in the Bases of NUREG-1431, Rev.1 upon i implementation of the new standard TS. In addition, schedule restraints of the worn l Attachment 21 39

i rcquired to movs irrtdi .ted full in ths vissil aft:r a shutdown prevent ths 100 hours from being cxce:ded: th r: fora, th3 scr:cning crit rion application question 2 is cor ectly answered with a "no".  ; Decay time is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either the failure of or presents a challenge to the integrity of a fission product barrier. Decay time does not satisfy criterion 3. { Decay time has not been shown to be significant to public health and safety by either operating experience or PRA. Decay time is not modeled in the CPSES IPE. , However, there is no indication that this function would be identified as risk significant if it was modeled in PRA models. Therefore, this TS does not satisfy criterion 4. (4) CONCLUSION _. This Technical Specification is retained.

               .X.      The Technical Specification may be relocated to a controlled document.

i 1 } Attachment 21 40 t -- - - - - - - - - - - - - _

T ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q3.9-7 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 3.9.4 c 1) Footnote " (Comanche Peak) CTS 3.9.4 c 1) Footnote ** and 4.9.4.1 Footnote " (Callaway) CTS 3.9.4.c and 4.9.4 Footnote * (Diablo Canyon) CTS 3.9.4.c Footnote " and 4.9.4 Footnote " (Wolf Creek) DOC 4-10-LS-20 iTS 3.9.4 NOTE and SR 3.9.4.1, JFD 3.9-11 In DOC 4-10-LS-20 and JFD 3.9-11, it was stated that this change is consistent with traveler WOG-76. Comment: Revise DOC by providing the TSTF number associated with WOG-76 and when the associated TSTF was approved. If WOG-76 has not made it to the TSTF process or the TSTF has not yet been approved, remove this item from the submittal since the inclusion of this footnote will be pending on the approval of the TSTF change. FLOG Response: WOG-76 was initiated by the WOG Mini-Group in October 1996. While we recognize that this is a generic change to the STS, the change was approved by the Westinghouse Owners Group over 18 months ago and was expected to have been approved by this time.' We expect the TSTF committee to complete their review of WOG-76 in the very near future. We believe the technical merits of the change are consistent with traveler TSTF-68, which should justify rapid approval by the NRC. This traveler is of sufficient value in precluding confusion, LERs, and inspection findings that should we be required to remove it from our submittal, an LAR would be submitted upon NRC approval of the TSTF. We believe that it would be cost effective for all concerned to retain this change within the submittal pending NRC review of the proposed traveler. Additional information supporting this DOC is in NSHC LS-8 in Enclosure 4. DOC 9-03-LS-8 is revised to include the following information:

          "A note is added to LCO 3.9.4.c and the 7-day SR to state that containment penetrations that provide direct access from the containment atmosphere to the outside atmosphere may be open under administrative controls. The note would allow these penetrations to be unisolated during CORE ALTERATIONS and movement of irradiated fuel assemblies within containment provided that specified administrative controls were employed. The proposed Note is acceptable based on administrative controls that consist of written procedures that require designated personnel having knowledge of the open status of the valves in question and specified persons designated and readily available to isolate the open penetration in the event of a fuel handling accident. These administrative controls provide protection equivalent to that afforded by the administrative controls used to establish containment closure for a containment personnel airlock. The NRC staff has allowed changes to the requirements for airlocks that allow both doors of an airlock to be open during CORE ALTERATIONS and during l          movement of irradiated fuel inside containment provided that administrative controls are i          in place to quickly close one door and establish containment closure.

The isolation valve, or temporary closure device, serves to limit the consequences of accidents. The proposed change would ensure the isolation valves, or functional equivalent, will perform their required containment closure function and will serve to litnit the consequences of a fuel handling accident as described in the Safety Analysis Report such that the results of the analyses in the Safety Analysis Report remain bounding. In considering the consequences of a design basis fuel handling accident inside containment, the assumptions in the analysis take no credit for the containment as a barrier to prevent the postulated release of radioactivity. For events that would occur during CORE ALTERATIONS or movement of irradiated fuel assemblies, containment closure is considered a defense-in-depth boundary to prevent uncontrolled release of radioactivity ~ For DCPP, a p eliminary dose calculation has been completed in accordance with the Reviewer's Note added by traveler WOG-76. This calculation shows sufficient time for closure with acceptable dose consequences. ATTACHED PAGES: Encl. 3A 5, 5a I f f 1 __ ______________ ____ _ __ _____ _m

CHANGE NUMBER MSBC DESCRIPTION i CVI signal would be generated by the containment gaseous  ! monitoring instrumentation and then verifying the valve I closure capability by remote manual closure from the control room. The combination of these two activities completely tested the automatic closure circuitry of l these valves. The NUREG 1431 surveillance requirement i performs the same function. The details of how the test is performed is not required to be in the TS.

                                                                                                              )

4 07 LG The specific administrative controls used to assure personnel airlock closure capability would be moved from J l the LC0 to the Bases. This change moves details that are ' l not required to be in the TS to protect the health and

safety of the public while retaining the basic limiting l conditions for operation.

4 08 LG Not applicable to CPSES. See conversion comparison table (enclosure 3B). 4 09 LS 14 LC0 3.9.4 would be modified to permit an approved functional equivalent of a valve or blind flange to isolate containment penetrations. This change is consistent with NUREG 1431. 1 4 10 LS 20 Adds a footnote stating that penetration flow paths that i provide direct access from the containment atmosphere to the outside atmosphere may be unisolatedAr adminis vt, controls,.JMnote is,addedjtoJLC0;3.9;4;C m 0 3.9 7 _ athe;7]daESRitpIstateithat' containment;penetrati.ons that1 provide!: direct:ar; cess;fros;the;~ containment atmosphereitoltheio$sideEateo. sphere'may;belopen;undet administrativecontf01sEThe. note;woul.dla]1owithese penetrations [to;be!Jnisolated during;COREfAL]ERATIONS end movementlofit rradt stedsfuellassemb]jeslwithjn! containment provided;thatispecYfied1 administrative controls 1were employedEghe proposed; Note 11siacceptable;basedion administrative: controls;that:consistlofiwritten procedures thatirequiteidesignatedjpersonnellhaving'knowledgefofAthe open;statusiotthelvalVeslinquest1onlandispecifjed personsidesignat_ediandfeedilylavailableitoliso]ateithe open:penetra_tionlinitheleventiofialfuel .; hand 11ngiaccidentG Jhese;administrativeicontrolsiprovidefprotection equivalentitolthat?:af forded ; by;the:admini strati ve: control s usedito;establishisontainmenticlosurelforla: containment personne12 air]ockEHhe;NRCistJffihas;allowedchangesito l theitequirementsiforlair]ocks;that:allowboth; doors:offan aitlockitolbe;openiduring;COREALTERATIONSlandduring movement;ofeirradiatedifuel?jnsjdeLcontainment provided thatladmin1 Frative ; cont @arelin! pl ace, to Lqui ckly: cl o.se eidoorlandles_tabli sh.; containment "cl osure! CPSES Description of Changes to CTS 3N.9 5 7/29/98 J

I l CHANGE NUMBER RSE DESCRIPTION i The11 solation valvekorctemporary;closureidevicel M0 3.9 7 h setvesitojjmitithe; consequences ofiaccidentsMThe mummmme a proposed; change would AnsureLthCJsolatjonlyalvesE or functionaEsquivalentRWil Kperforalthetrlrequired containment;closurejfunction andMlliserveit63init;the consequences;of alfuel; handling a_ccident:asIdescribed;in the LSafety; Analysis;Reportisuchlthat(the: resul ts!ofuthe j analyses 11nLthe; Safety lAnalysisReport"remainboundingij; L Iniconsiderjngithe: consequences 1of;a idesign; basis! fuel l handlinglaccidentiinsidefcontainmentEthe; assumptions;in the:analysisitake;no credit;fotthelcontainmentJas:a ) l i barriet;to weid.ithelpostul atedirelease;of radioactivity 2 ForleventsLthatlwouldloccuriduring; CORE ALTERATIONS orimovement;ofiirradiated;fue1[ assemblies; , containmenticlosureli s; consideredialdefense e in!; depth ! boundary'to' prevent uncontrolled; release lof' radioactivity. V I This change is consistent witn traveler WOG 76 and with previously approved administrative controls for personnel air locks. 5-01 R This change relocates the current TS section dealing with maintaining direct communication between the control room and the refueling station to a licensee controlled document as part of the conversion of the current Technical Specification to the format and expanded Bases 0 3.9 9 of the improved Standard Technical Specifications. The specification;requiresicommunication;between!the; control roonland!thelrefueling stationito: ensure]that:any; abnormal. change 11nitheifacility status;orJcore[reactivityiobserved onitheicontrol" roomlinstrumentationican; be; communicated Lto the refueling! station:personneliduringicore alterations? I l 1 Ihts: proposed;TS revision; relocates; requirement:Rwhichido ' i not(meetithe1TS;critetiain,10CFR50.36(c)(2)(11)Rto documents 1with established controllprogramsEThis regulationladdressesitheiscope2and^purposelofiTS E In doing soEitisets2forthlaispecific;setiofcobjective criteria;foridetermining;whichiregulatory; requirements and operating; restrictions should;beltncluded21.nithelTS7 Relocation 1offtheserequjrements;allowsjthe;TSitoibe resetved ~;onlEfor;those" conditions;onl imitations;upon reactor;operationwhich;are1necessaryfto_ obviate 3he l possibility;offantabnormalisituationior; event lgivingirise tofan;jamediatelthreatito mthe:publ.ic9ealthland; safety therebylfocusing; thei scopelofs thelTS EAnTeval uationiofithe applicabil itylofdtheseicriterialtolthi s!speci fication "i s ProvidedliniAttachment?21; Tolensurelaniappropriateileve1LoCcontrolEthese requirements;w11Rhe relocatedito:1): documents;thatlare subject 1to_the: provisions 1of;10;CFR;50.59,12)1other CPSESDescription ofChanges to CTS 3M.9 Sa 72 9/98

i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-8 APPLICABILITY: CA, CP, DC, WC i REQUEST: CTS 4.9.4 a (Diablo Canyon and Wolf Creek) CTS 4.9.4 a.1 (Comanche Peak) CTS 4.9.4.1 (Callaway) DOC 4-03-LS-5 l ITS SR 3.9.4.2 - The frequency to verify the occurrence of containment ventilation isolation is proposed to be changed from 7 days to 18 months. Other than the statement that this change is consistent with NUREG-1431, Rev.1, the DOC does not address any specific technical justi;iations associated with this change. Comment: Revise the DOC to include specific technical justifications for this change. FLOG Response: DOC 4-03-LS-5 has been revised to include specific technical justifications for this change. ATTACHED PAGES: Enci 3A 4a I

CHANGE NUMBER EitE DESCRIPTION , required to be in the TS to protect the health and safety of the public while retaining the basic limiting conditions for operation. 4 02 LS-4 Removes Surveillance requirement to perform verification within 100 hours prior to the start of core alteration or movement of irradiated fuel. This is consistent with l NUREG 1431. Rev.1, and is acceptable because the deleted i requirement is redundant with the requirement to meet the ! LC0 at the time that CORE ALTERATIONS or fuel movement begins. l 4-03 LS-5 The frequency of verifying that [ containment ventilation l isolation] occurs is changed from 7 days to 18 months. This is consistent with NUREG-1431. Rev. 1. This change is acceptable because the revised frequency requirement will continue to assure the OPERABILITY of the valves. The, proposed; change;to ,181monthstwoul.d ; apply;the:same 0 3.9 8 frequency;of;testingito: containment; purge;1 sol ation yal_veslasii slappl iedito;other.Leontainment(i sol ati.on valves,that;must;be10PERABLEidurjng reactor; operations.] l The118; month ifrequency;ha s; been; found "adequatelfortthe typeiofitestingLappliedit61nstrupentationf andf valvesithat must; mitigate events 1muchimorelsever,eiandinuch more l challenging;tolthe; containment;boundaryf(eig.t.LOCA7MSLB). than3hefHA] The new frequency is consistent with those ( SRs applicable to ESFAS type functions and inservice valve testing which are appropriate for the containment isolation function. 4 04 TR 1 Revised Surveillance requirement to allow for increased flexibility in using an actual or simulated actuation l signal. Identification of the specific signal is moved i to the Bases. 4 05 A Not applicable to CPSES. See conversion comparison table ) (enclosure 3B). i 4-06 LS-23 The requirement to verify the capability to close the containment ventilation isolation valves from the control room would be deleted. The LC0 requirement is to have operable automatic containment ventilation isolation i valves. The CTS surveillance to verify this capability l was poorly written in that it demonstrated the automatic I functioning of the valves by first demonstrating that a CPSES Description of Changes to CTS 3N.9 7/29/98

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-9 APPLICABILITY: CP, DC i REQUEST: CTS 3/4 9.5 DOC G-01-R CTS 3/4.9.5 is proposed to be relocated to an unspecified licensee controlled document. The DOC does not provide any technicaljustification supporting this relocation. I Comment: Revise the DOC by providing additional justification for the relocation and identify I the licensee controlled document containing this requirement. This requirement shall be } relocated to a licensee controlled document controlled by 10 CFR 50.59. FLOG Response: DOC 5-01-R has been revised and a Technical Specification Screening Form for CTS 3.9.5 has been prepared to provide justification as to why the relocation is acceptable. This justification shows that this LCO provides for the ability to communicate any abnormal changes to the fac"ity or core reactivity to the refueling bridge during CORE ALTERATIONS. This communication is not applicable to any instrumentation used to detect a significant abnormal condition. It does not apply to any initial condition of a DBA or transient analysis. It is not a part of the primary success path in mitigating a DBA or a transient. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch l reviewers during telephone calls on June 25 and June 30,1998. The document indicated in l Enclosure 3B is controlled by 10CFR50.59. l ATTACHED PAGES . 1 Enci 3A Sa & Sb Att 21 41 & 42 l

CHANGE NUMBER HSBC DESCRIPTION The21solatjortvalvelCor:; temporary; closure; device? 0 3.9 7 serves 1to11joitlthe1 consequences;ofiaccidents E Ihe Proposedichangeiwouldlensureathe31so]ationivalvesEor functionallequivalentswilliperformitheitirequired containment: closure 1 function .andlwilliserveitollikittthe consequences;oftalfuel;handl ing l accident;asidesetibed Lin the; Safety:AnalysisJeportisuchj thatithe; resultslofithe analysesijn;the; Safety AnalysislRepott1 remain; bounding;j In;considerjngittucconsequencesioffaldesign; basis 1 fuel handijng; accidents;1nside: containment 7;the assumptions;1n the: analysis;take;norcredittforithelcontainment' ass a bartier;toJ preventEthe;postul ated[teleasej of radioactivitygfordevents;that Would;occurZduring CORE ALTERATIONS;orimovementlof;it radiated ;fuellassembl ies ? containment;closuresistconsideredaidefensedin; depth boundaryj to: preventi uncontrol lediteleas_elofiradioactivity; This change is consistent with traveler WOG 76 and with previously approved administrative controls for personnel air locks. 5 01 R This change relocates the current TS section dealing with maintaining direct communication between the control room and the refueling station to a licensee controlled document as part of the conversion of the current Technical Specification to the format and expanded Bases 0 3.9 9 of the improved Standard Technical Soecificat . l GEsanon ayydie's[communicationlbetweenithe: control room;andithelrefueling station;toiensuretthatlanyl abnormal change:jn;the;faciljtyistatusiorigoref reactivitylobserved on1the:contro]ZroonJjnstrumentation:canibelcommunicatedito thelrefueljng; station personne11duringicore~ alterations j Thi sipropo.sedlTSrrevj sion: relocates; requirements;iwhichido notimeetitheJTS criterjalin 10CFR5013.6(c)(2)(ii)Eto documentsiWithlestab11shed controllprogransRThis 4 regu]atjoniaddtessesithe:scopelandipurposeiof3Sl2In l doing;sonitisettnfotthiaispecific set!of! objective  ! criteria?for; determining [which; regulatory;requirementsiand operating:restrictionXshould; beiincliadedlinitbegTS! Belocation of;the_sefrequirements3110wCtheiTS;toibe reservedionlylfotithosefconditionslorsljaitationslupon teactor;operationlwhich are;necessaryito; obviate;the possibil j ty;off.anj abnormal;situationierJevent ;giving ;rj se tolan11mmediateithreat3olthejpub]jeLhealthiand safety therebylfocusjngithelscope ofjthe1TSCAn; evaluation offthe app]icabiltty"ofitheseictiteriato;this1specificationfis PCovidedlinMttachmenti217 TolensurelanLappropejate;1evelioficontro1Mthes_e requirementslwillibe; rel ocated ito:1) EdocumentsJ thatt are subjectitolthe! Provisions;ofc10 CFRs 50;59K 2)[other CPSESDescription of Changes to CTS 3M.9 Sa 7/29/98

CHANGE NUMBER MtiC DESCRIPTION

                                             . documentslwhichlhave(similardregblatory 0 3.9 9 controls:(e_;g g the;Qua11ty4 Assurance; Plan Ras described;initheiFSARnwhichliscontrolledjby 10CFR50;54a) Gor!3);to; programs 1that;are; controlled,via the: Administrative 1 Control s ;sectionlof;the; improved 1TSI:i             {

The11dentificatjonlof;the specific lilicenseeicontrolled  ! document containingithis;rcuirement:1siprovidedl;in l

Enclosure:

3Blofithejconversionssubmittalf I i Compliance;withithe! relocated l requirements;willinottbe , affectedlbylthi.s;proposedichangeito;the; current; Technical

                                                                                                                 ]

Specif,1 cations M hel. required;periodicisurve111ances'w11], . contjouetojbepetfotmedjtojensureithat31mits;on s  : Parameterslate maintained E Therefore grelocationiof;these l l' requirementsiwillihaveinolimpaction; system operabilityfor the: maintenancelof2controll ed; parameters;withinf l imit sa , 6 01 R This change relocates the current TS section for the Refueling Machine to e licensee controlled document as part of the conversion of the current Technical Specification to the format and expanded Bases of the improved Standard Technical Specifications. The 0 3.9 10 OPERABILI K requirementsifor;thelrefueljng machine main; hoist:andyauxiliary:monorai11 hoist; ensure 1that (1);the mainihoistiw1111be:used;forimovementfofffuel as.semb11esa(2)1the: aux 111ary monora1Rhoist:will!be used fot;1atching Eunl atchinglandinovamentio fl contro11 rod :dri ve , shaftsE(3)1the ma1Lhoist;hasisufftcient[loadicapacity;to lift:a;fustassembly1(withicontrollrods)R(4);the aux 11taryamonota111 hoist:has: sufficient;1oad capacity;to l atch gunl atchjandinovelthe icontrol " rodidrivelshafts'i;and (5);thelcore11nternals andireactor.(vesselfarel protected ftomiexcessivellifting1 force;in;thefeventithey;are inadvertent 1Eengagedf.durjng111fting operations? Thisjproposed1TS;revisjonfrelocates'jrequirementsEwhichido < l notiseetithe;75'ctiteriaijn110CFR50;36(c)(2)(11)Kto documents 1withl establ i shed ; cont _rollprogransEThi s regulation address.esithe" scope;and;purposelof;TSljIn doing(sogjtisetsjforth'aispecific, set;ofl objective etiterialfot; determining whichiregulatory; requirements;and operating; restrictions j shouldlbe sjncl udedli nithe;TS R Relocation of}these; requirement.slallowsithe;TStto:be reserved;on]ymfot;those conditionsforglimitationslupon teactor; operation;WhichiarejnecessaryLtolobviat.ejthe possibilitylof :anTabnormali situationiotievent;giving ;ri se to aniimmediate:threatito;the;publicihealthlandi. safety thereby1 focus.ing_the;scopepfihe;TS.IAn; t evaluation;ofithe appljcabiljty of;theseicriterialto;this specificationiis provjdedjinl

Attachment:

21; CPSES Description of Changes to CTS 3/4. 9 Sb 7/29/98 i

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.9.5 COMMUNICATIONS Applicable MODES: During CORE ALTERATIONS (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: YES NO _ X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ X (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ X. (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION This specification requires communication between the control room and the refueling bridge to ensure that any abnormal change in the facility status or core reactivity observed on the c'antrol room instrumentation can be communicated to the refueling bridge personnel during core alterations. The TS requirements for communications are not applicable to installed I instrumentation used to detect a significant abnormal degradation of the RCPB; , therefore, this TS does not satisfy criterion 1. i The cornmunications TS is not associated wi*h a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challonge to the integrity of a fission product barrier. Thus, t'1is requirement does not meet criterion 2. The TS for refueling communications does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the requirements do not satisfy criterion 3. Communications during core alterations has not been shown to be significant to public health and safety by either operating experience or PRA. Communications during core Attachment 21 4i L_____--_-----------

altarttions is not mod: led in th3 CPSES IPE. Howev:r, thIra is no indication that this ! function would bs idrntified as risk significant if it was mod:ltd in PRA models. ' T!.orefore, this TS does not satisfy criterion 4. (4) . CONCLUSION _ This Technical Specification is retained.

                ,2L      The Technical Specification may be relocated to a licensee controlled docurnent.

1 i I l Attachment 21 42

ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: Q3.9-10 APPLICABILITY: CP, DC REQUEST: CTS 3/4.9.6 DOC 6-01-R CTS 3/4.9.6 would be entirely relocated to an unspecified licensee controlled document. The DOC does not have sufficient justification to support the relocation. Comment: Provide additionaljustification as to why this relocation is acceptable and identify the name of the licensee controlled document containing this requirement. This requirement sha9 be relocated to a licensee controlled document controlled by 10 CFR 50.59. FLOG Response: DOC 6-01-R has been revised and a Technical Specification Screening Form for CTS 3.9.6 has been prepared to provide justification as to why the relocation is acceptable. This justification shows that this LCO provides the ability to lift and manipulate fuel assemblies. It is not associated with installed instrumentation used to detect a significant abnormal condition in the reactor coolant pressure boundary. It is not associated with any variable or condition which is the initial condition of a DBA or transient which may challenge a fission product boundary. The manipulator crane is not a part of the primary success path for , mitigating a DBA or transient. The format for specifying the location of relocated requirements (in Enclosure 3B of the i conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998. The document indicated in Enclosure 3B is controlled by 10CFR50.59. ATTACHED PAGES: Enci 3A Sb & Sc Att 21 43 & 44 1 I l l j i I

CHANGE NUMBER RSliC DESCRIPTION licenseeldocuments;which;have sim))at!tegulatory 0-3.9 9 control s ,(e . g R the _ Quality ~ Assurance: P]an Ea s l descrjbed;inathe:fSARDdlichli sIcontrollediby 10CFR50iS4a)Zer;3);to;prograssithatIare;contro11ediyja the: Administratjye1ControlsJsection;ofithejjaproyed; TSE Thel 1dentificatio[oflthe] specific 11]censee; controlled document:containj ngithis: requirement 11s1provided ;1n Enclosure;38;ofLthe*conyersion;submitta]; Compljance;with the relocated, requirements;willnot;be affected by;this~proposedichange:tolthe; current; Technical SpecificationsMThe requjred periodicisurveil]ances;will continuem to,be:performeditoiensureithatilimitsion parameters arelmaintajnedleTherefore;;telocation:of ithese requirements;wilijhave nofimpact;onisystemloperability or the; maintenance offcontrolledl parameters.withinflimitsO l 6 01 R This change relocates the current TS section for the Refueling Machine to a licensee controlled document as part of the conversion of the current Technical Specification improved Standardto the format Technical and expanded Specification I The- Bases 0 3 910 ' of[l ILITYLrequirements; Tor _.y m,rerueling machine mainihoistiand auxiliarfmonora11thojstiensure:that: (1)ithe;mainthoistiwil_l;be;usedifor;novementlof fuel assemb] iesR(2)lthelaux111ary _ monorail; hoi stiw1111be_ used , forilatching,iunlatchinglandinovementx ofccontrol. rod driye shafts',l:^;(3);the; main 1 hoist;pasisufficientJoadicapapity;to li ftla2 fuel;;as.sembly;(with;pontro] Erods)/;(4)1the auxiliatfmonoraillhoistihasisuffj'cient:1oadicapacity;to 1.atchEunlatchtand moveithe; control; rod;drivelshafts;Tand (5)ithe: core]j nternal slandirea.ctotg vesseliarej protected fromlexces.sj vellj ftjng lforcelinitheleventlthef are inadvettant]ylengaged;durj ng ilj fting! operations: This;proposedlT5!reyision;relocatesfrequirements,1which;do not:me$thelTS1critettalin110CfR5013McM21(11) Cito documentsLwitgestablished controliprgtansHThis  ! I regulationytessesithelscope3nd; purpose;ofiTS;EIn doing;soaitisets;forth!alspecificisetioLobjectiye etiteriaroridetermining' which;regul atom requirements:and operating;re.strictionsishouldibelincluded;jnithe TS; Rel ocatioglofithes.elrequirements"allowsithe;TSitol be reserved;onlyfor2thoseiconditionslori] imitations.~upon reactorloperationiwhichfare:nece.ssary to obviate.the possibilitylofian;abnormaEsituatioq;orleyent;giving rise to;anlimmediate, threat,.to;thel publj g theal th:and; safety therebyffocusingithe;sc. ope lof;thejTS!sAnlevaluationlot3 t applicabilitylofithese crjteria';tolthis specificati s provided[in' Attachment l21; { l 1 CPSES Description of Changes to CTS 3N.9 Sb 7/29/98 J

CHANGE NUMBER NSliC DESCRIPTION To;ensurelan tapproprj ate 11eyeKoficontroltthese 0 3.9 10 requirementslwill;be; relocated;to;1); documents;that are:subjectatolthef proyj sions;;ofl10;CFR L50; 5922) otheriljcensee; documents!which; havelsimilarl regul atory controlsI(e;ggtthe;Qualjty;Assutance P.lanEas described igithe;FSAREwhichlisicontrolledibyIl0CER50',54a)2orJ3)ltc { programs 1thatiareicontrolledfyia;thelAdministrative Control s;section;pfithelieproved hTSslfTheitdenti fication ofitheIspeciffc1]icens.eelcontrolled:documenticontaining this1te4jrementIisiprovided;1n1 Enclosure 13Blofithe conver.sionisubmittalg Compljance:withithelrelocatedirequirements;willjnotibe affected;byltMsJ ptoposedichangelto;the;currentJechnical SpecificationsMThelrequjred perjodicisurveillances;will continuelto;beiperformedito;ensureithat;]imits;og Parameters;areimaintained;dThereforeErelocationiofithese requirements;wil1 ~ havelnolimpaction; system;operab111ty1ot

                                       ;maintenanceloficontrolledipataseters;withiniljaits? j 7 01            R             This change relocates the TS section dealing with                   0 3.9 11 crane travel to a licensee controlled document as part of the conversion of the current Technical Specification to the format and expanded Bases of the improved Standard Technical Specifications. The restriction;on; movement"ofjoadsliniexcessjofitheinominal weight;ofJ a[fue));andlcontrolf rodlassemblyla_ndlassociated hand 1j pgitoollover;other[ fuel ;a.ssembl iesiinlaistorage; pool ensures 2thatsin1the;eyentithisiloadlisidropped:M1)~ the activityReleaselwilEbellimiteditolthaticontained;jn a singlefj]fassemblyEa!)dE(2) Zany possjbleldistortjonlof fue13jn;thelstorageiracks;W111;notIresult;inJa;crjtical arraye2This!assumptjon11s1 consistent with thejactiyJty tele se assumed;1Etheisafetytanalyses; This; proposed1 TSirevisionLrelocatesErequirementsawhich1do notimeet;the TS;criterialirH10CFR50;36(c)_(2).(11)Zto documentsvithlestablished contro11 programs;dThis regul ationladdressesithe 'scopeland; purpose;of2TS gin doingsogit: sets;forthia!specificisetioffobjective crjterialfor;determiningwhichiregulatoryirequjrementsland operating fes.trictionsishoul d[beij ncl udedlinitheJSI Relocation;of;these; requirements 2 allows (the;TS;to be reserved lonly;for;} hose:conditionsfor;1jmitations;upon

• reactorloperation;whichlare_necessary;toiobyJateithe possibility l.oflanlabnormalisituationf orlevent giyingirise to?anij amediate tthreatitolthelpublic; heal th ;and ; safety therebygfocusingithelscope:ofithelTS CAn!eyal u6tioniofJthe applj cab 111tylof1these ;crj teti alto tthi s;speci fication;i s

, provjdedljri

Attachment:

(217 i CPSES Description of Changes to CTS 3M.9 Sc 7M988

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.9.6 REFUELING MACHINE Applicable MODES: During movement of fuel assemblies and/or latching, unlatching or movement of con'.rol rod drive shafts within the reactor vessel (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis - that either assumes the failure of or presents a challenge to the integrity of e fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of the above questions is *NO", the TS may be relocated to a controlled document. l (3) DISCUSSION { The requirements for the refueling machine main hoist and auxiliary monorail hoist ensure j that: (1) the main hoist will be used for movement of fuel assemblies, (2) the auxiliary j monorail hoist will be used for latching, unlatching and movement of control rod drive shafts, (3) the main hoist has sufficient load capacity to lift a fuel assembly (with control rods), (4) the auxiliary monorail hoist has sufficient load capacity to latch, unlatch and move the control rod drive shafts, and (5) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations. i The TS requirements for the refueling machine are not applicable to installed l h strumentation used to detect a significant abnormal degradation of the RCPB; ! therefore, this TS does not satisfy criterion 1. The refueling machine TS is not associated with a process variable, design feature, or operating restriction that is monitored and controlled and is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the l integrity of a fission product barrier. Thus, this requirement does not meet criterion 2. l The TS for the refueling machine does not app!y to an SSC that is part of the primary I Attachment 21 43 j

l l-success path and which functions or actuat:s to mitigats a DBA or tr:nsi:nt that cithar cssum:s ths failurs of or pr:s:nts a challenge to tha integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3. The requirements of this technical specification are not a significant risk contributor to public health and safety by either operational experience or PSA. The refueling rnachine is used to transport fuel assemblies during refueling operations. The CPSES IPE models the plant during power operations, and therefore does not include the refueling machine in any risk quantifications. However, if the refueling machine were included in the me del, its significance would be negligible. Therefore, these requirements do not satisfy criterion 4. (4) CONCLUSION _ This Technical Specification is retained.

                .)L      The Technical Specification may be relocated to a licensee controlled document.

i i j l I l l l Attachment 21 44

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-11 APPLICABILITY: CP,DC REQUEST: CTS 3/4.9.7 l DOC 7-01-R CTS 3/4.9.7 is proposed to be relocated to an unspecified licensee controlled document. The DOC does not provide any technicaljustification supporting this relocation. Comment: Provide additional justification as to why this relocation is acceptable and identify [ the name of the licensee controlled document containing this requirement. This requirement shall be relocated to a licensee controlled document controlled by 10 CFR 50.59. FLOG Response: DOC 7-01-R has been revised and a Technical Specification Screening Form for CTS 3.9.7 has been prepared to provide justification as to why the relocation is acceptable. This justification shown that the fuel handling building crane travel is used to restrict loads in excess of a fuel assembly from traveling over the spent fuel pool. It not associated with any instrumentation used to detect a significant degradation of the reactor , coolant pressure boundary. While it does provide a restriction used to prevent a heavy load j drop event, it does not apply to initial condition of a DBA or transient analysis. Crane travel l over the spent fuel pool does not provide a primary part of the primary success path for mitigating a DBA or transient. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch reviewers during telephone calls on June 25 and June 30,1998. The document indicated in Enclosure 3B is controlled by 10CFR50.59. ATTACHED PAGES: Encl 3A Sc & Sd Att 21 45 & 46 l i l L )

CHANGE NUMBER lEllC DESCRIPTION To;ensu.relan;approprj atejleveliofJcontrolElthese n 3,9,1o requirements 31]Ebe: relocateditoI1Edocumentsithat areZsubjectitolthe provisions 1of;101CFBiS0;59E2) otherg11censeeldocumentsWhjchihaveltsinRar; regulatory controls 1(Kg;EtheiQuality3ssurancellanHasidesctibed j nithelFSAR,1whichitsi controlled ' by21.0CFR50154a) EorJ3);to programsithat[arelcontrolledivtalthe: Administrative Controlsz_sectionrofithe;1mproved3SriThendentifjestion ofithelspecific;]Jcenswicontrol)ed:documenticontaining thisJtequirement318;pfoy1dedyjn

Enclosure:

381of;the conversionisubmitta11:1 Compliance;with:thelrelocated requirements;;willWAbe affected;byithis proposed:changeitolthelcurrentRechnjcal SpecificationsMThe1 required :perj odicisutveil)ances:Mll conti g tolbe performedi tolensureithat;]imitsjon parameters arelaatstained:aThereforegrelocationlof;these requirements 1wn]Ihavelno;japactionisystenloperabilityaog the~maintenanceioficontrollediparameters Withinilimits? 7 01 R This change relocates the TS section dealing with [ 0 3.9 11 crane travel to a licensee controlled document as par of the conversion of the current Technical Specification tM femat and =psded o es of the impre tandard Technical Specifications. The rj etionloninovamentj ofiloadsij niexcessiof;the1 nominal

                                       .ightlofielfue1Eandcontro11rodiassemblyland" associated handijng;tootoyeriotherifue1EassembliesiinlaIstorage: pool.

ensures;thatlinitheleventithisIload11sidroppedC(1) the actlyityireleaseMlEbe111miteditolthaticontainedlin;a singlelfuelrassemblygand1(2):any;pos_sibleldistortion;of fueMin;thelstorageltacks;wi1Enotitesultjin;alcr(t1 cal, attayr2Thislassumption11siconsistentyithithelectivity releaselassumedlinithelsafetyianalysesi Thi si proposed:TS] revision :relocatesi requirements uwhich'do notisset;the35;ctitetjalipi10CFR50 36(c)(21(11)f,Sto documentsiwith:establishedicontro11progransEIhis regulatigaddresseslithe: scope [and; purpose lofLTSZIn doing;soultisetsiforthlaispecificiset;offobjective  ; ctiterialfotidetermining;whichiregulatoryrequirements:and i I operating;testrictionsishpuldibelincludeddin;the1TS7 l RelocationlofitheseIroquirementslallowsitheiTSito; be j reseryed:pglylforlthose; conditions;orllimitationsiupon { reactorioperatio O hichiare necessaryitoiobviateithe ' possib111tyiofEahlabnormallsituation ;orj eventigivingl rise tOpliamedj atelthreatl tolthelpublic1 heal th;and: safety therebylfocusingithelscopelofithe;TSTIAnievaluationiofit l applicability;of;theselcriteriaito;this; specification;1 rov,1dedlinAttachment2217 I l 4 CPSES Description of Changes to CTS 3N.9 Sc 7/29/98 W-_ _- -

CHANGE NUMBER R$HC DESCRIPTION . ! o'ensurelaniappropriate:1evello([ control;?;these 9,3,g,11 te@irementsMthe;telocated; toll);documentsitha t areIsub M it 0 the;provjsions o E10lCFR150J59?i2). othetflicenseeldocuments W ehjhaveisimilar1 regulatory controls 3esg;2tW; Quality; Assurance;P,lan;;as;desetibed jn;thef fSARDmich;1s; controlled:by30CF,R50:54a) Eor;3)1to progtassithatlare;controlledivialthe~ Administrative

Control s; sectionlof;the11mprovediTS MJhelj identification of;thelspecificRicenseelcontro11ed;doeurent1 containing thistre@jresentgisrprovidedlin! Enclosure 138;;of;the conyersionisubmittalq comp]1ance

with thelrelocateditequirementsyt1Enotibe affected;by # is;proposedichange:tolthe;cutrentiTechnical Speci fjcations Fj;;The;; required; perjodic;sutve11)ances1w11]

continuelto;be; performed!to;ensureithat"Jimitsion parameters are;maintainedmThereforetrelocation;of;t se retirements l311thayeino;:impactionlsystenloperabiljt ot

hejmajntenanceloficontrolled

parametersiwithin:11 s

_7

                                                                                           %                                                          \

1 l l l CPSES Description of Changes to CTS 3M.9 54 7/29,yg

                                                                                                                                  ._________-________a

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS Applicable MODES: With fuel assemblies in a storage pool (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. _ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. I _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that ei+.her assumes the failure of or presents a challenge to the integrity of a fisslun product barrier. _ 1 (4) An SSC which operating experience or probabilistic safety assessment (PSA) has shown to be significant to public health and safety. If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all four of the above questions is "NO", the TS may be relocated to a controlled document. (3) DISCUSSION This specification ensures that loads in excess of one fuel assembly containing a control rod, plus the weight of the fuel handling tool, will not be moved over other fuel assemblies stored in the spent fuel storage racks. Therefore, in the event of a drop of this load, the activity released is limited to that contained in one fuel assembly. This also prevents any possible distortion of fuel assemblies in the storage racks from achieving a critical configuration. This specification applies to prevention of a heavy load drop accident and assures that the damage caused by the load is hited to the equivalent of one spent fuel assembly. This assumption is consistent with the activity release assumed in the DBA safety analyses for a fuel handling accident. The TS requirements for crane travel are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1. l The fuel handling building crane travel TS is associated with an operating restriction for a heavy load drop event. This specification is not applicable to a process variable, design feature, or operating restriction that is monitored rnd controlled during power operation and is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this requirement does not meet criterion 2. This conclusion is consistent with the corresponding evaluation in Ref. 4. Attachment 21 45

i The TS for crans travtl does not cpply to an SSC that is part of th3 primary succ:ss I path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. { Therefore, these requirements do not satisfy criterion 3. From Ref. 3, the fuel handling building crane has not been shown to be a significant risk contributor to public health and safety by either operational experience or PSA. 1 Ref. 3 reviewed several environmental reports related to these cranes, and found their risk significance to be minimal. The spent fuel storage facility crane is not modeled in the CPSES IPE. Therefore, these requirements do not satisfy criterion 4. (4) CONCLUSION

                                                            ._       This Technical Specification is retained.
                                                            .X.      The Technical Specification may be relocated to a licensee controlled document.

l l l 1 l l l l. Attachment 21 46 l

  - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - ___                                   . _ - _ _    ._    -            _                      -     ---  -  A

ADCITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-12 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 3.9.8.1 Footnote * (Comanche Peak and Callaway) CTS 3.9.8.1 Footnote

• and ** (Diablo Canyon)

CTS 4.9.8.1 Footnote * (Wolf Creek) DOC 8-03-LS-6 The CTS requirement, which allows RHR loop to be removed from operation during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot leg, is proposed to be deleted. The DOC does not address any technicaljustification but states that this change would allow increased flexibility for core mapping and isolation valve testing, and that this change is consistent with NUREG-1431, Rev.1. In addition, there is not any discussions on the possible increase in risk associated with decay heat removal. Comment: Revise the DOC by providing the justification as to why this deletion is acceptable and how it relates to the current licensing and design bases. Was there any risk assessment performed in regards to this issue? If so, what were the conclusions that would support the proposed change? FLOG Response: DOC 8-03-LS-6 is revised to provide the following additional justification as to why the deletion of the restriction on the 1 hour per 8-hour removal of operation of the RHR loop is acceptable: "The proposed change would permit the securing of RHR flow through the reactor vessel for up to 1 hour in every 8 hours [] provided that no operations involving a reduction in boron concentration were performed. The current TS already permit these interruptions in RHR operation but under more limited conditions. During this interruption, decay heat removal is assured by natural convection within the large amount of water in the refueling cavity (2 23 feet above the reactor vessel flange). Boron concentration concerns are avoided by prohibiting evolutions that would reduce the boron concentration. During MODE 6 operation with water levels 2 23 feet above the reactor vessel flange, the reactor coolant temperature is maintained significantly below boiling with the typical time-to-boil significantly in excess of [1 hour). Decay heat removalis provided and inadvertent criticality is avoided." ATTACHED PAGES: Enc! 3A 6

CHANGE 1 l NUMBER EliC DESCRIPTION i 8 01 A This change, consistent with NUREG 1431, provides technical guidance and clarification that loading irradiated fuel assemblies in the core is the specific activity of concern that could increase the reactor decay heat load. This is not a technical change because, under these conditions, the only activity that could increase , reactor decay heat load is loading irradiated fuel into j the reactor vessel. 8 02 A Not applicable to CPSES. See conversion comparison table (enclosure 3B). 8 03 LS 6 This change allows the removal of the RHR loop from operation for additional purposes other than the 3 p.:rformance of core alternations in the vicinity of the l hoc legs. This allows increased flexibility for core { mapping and isolation valve testing. No operations are 1 permitted that would cause a reduction of the RCS boron concentration' Ms-changa is_ consistent with NUREG-143 1. _he;proposedichangec woula permit 1the 0-3.9 12

                                   ._ecuting;of:RHRPow;through;the reactorives.seEforoup                           !

toll; hour;jnlevery 8: hours.f]!provided.;thatino j OPer_ationsfjnvolvingla reductjon;;in boron; concentration were1per formed y Thefcurrent JSTal ready 2 permit;thest interruptions;1n1RH.RoperationibutlunderzmoreMinited conditionsG!)urj ng ithi sl i nterruptionKdecay; heattremoyal j i slassuredj bytnatural i convectj on LWithin ;the il arge; amount  ! ofLwatetin3the3refuelingcavityi(g!23Meetlabovetthe reactor; vessel; flange) L Boroniconcentrationiconcernsllare avoided, by; prohibiti ng;elvol utionsi that/ would i reducelthe boron;.concentratjoni7 DuringLH0DE;61 operation yith watet l evel:1235 feet rabove ;thelreactor3essel ;fl angellthe reactoricoolantitemperature 1simaintainedis. significantly , below1bo11]ng with:theitypicalitimeito b91 Significantly j' in;excesslofi[lihour] U Decay;he.atiremoyal M s;providedland inadvertent icritical i tyli s Tavoided. l 8 04 A This enange eliminates the option of securing RHR prior to  ; initial criticality since initial criticality has I already occurred. Thjs;notelisinojonger applicable o.3,9 13 l for; pl ant: operations;andfi.siobsol etej cycleidependent  ! 1.nformationCJhi.shtypelofiinformationLeay;be removed underjanjadministrative(A),changelsincejnoAechnical s changesf(eitherJactual or interpretational)?are;being l made? 8 05 Not used. 8 06 M This change adds an additional surveillance requirement to verify correct breaker alignment and indicated power CPSES Description of Changes to CTS 3M.9 6 7/29/98 __-__________a

1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-13 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 3.9.8.2 Footnote

• DOC 8-04-A The CTS footnote regarding the option of securing PHR prior to initial criticality is proposed to be deleted entirely. This change is acceptable because it is a relaxation provided in the guidance of NUREG-1431, Rev. '. However, the categorization is in error. This change is not an administrative change, but it is a more restrictive change.

C'mment: Provide a ret ' red "L" DOC. FLOG Response: This note is applicable to " initial criticality" only. All start ups including those occurring after refueling outages may not apply this note since initial criticality has already occurred. This note is no longer applicable for plant operations and is obsolete cycle dependent information. This type of information may be removed under an administrative (A) change since no technica! :,hanges (either actual or interpretational) are being made. DOC 08-04-A will be revised to add the following statement; " This note is no longer applicable for plant operations and is obsolete cycle dependent information. This type of information may be removed under an administrative (A) change since no technical changes (either actual or interpretational) are being made." l ATTACHED PAGES: Enci 3A 6 l l l I

CHANGE NUMBR lEC DESCRIPTION 8 01 A This change, consistent with NUREG 1431, provides technical guidance and clarification that loading i irradiated fuel assemblies in the core is the specific ) attivity of concern that could increase the reactor decay heat lor.d. This is not a technical change because, under these conditions, the only activity that could increase reactor decay heat load is loading irradiated fuel into the reactor vessel. 8-02 A Not applicable to CPSES. See conversion comparison table l (enclosure 3B).  ! 8 03 LS-6 This change allows the removal of the RHR loop from operation for additional purposes other than the performance of core alternations in the vicinity of the hot legs. This allows increased flexibility for core mapping and isolation valve testing. No operations are permitted that would cause a reduction of the RCS boron concentration. This change is consistent with NUREG- , 1431. Rev. 1. TheJprcposed; change would permit;the securing;ofRHR;f).ow:through.;thelteactorvesse1;for;up s QM I l tEone hourfjn;every.;eightlhoursfprovided[no:operatipns involvingtaireductioniin1 boron 1 concentration were  ! petformedT3harcurrent;TS?alNadyJpermitithishinterruption jn:RHR: operation 1but;under; moreJ11mitediconditionsU The twoiRHRl functions areiadequatelylassuredibecause Edecay heatyemova.Ewillibe; accomplished byftoe: natural convectionitolthellarge amountJof;wateriinfandfabove;the 8C5f(G32feetLabove;the2eactorivessellflange), and, boron mixing concermate} avoided;bylprohit* ting; evolutions.that would:reducelthe.; boron: concentration.2The. on.e:out off eightLhourirestr,1ctioniis'adeqsate:tolensure proper: decay heatiremovalEDuring M00E1.6 operationsWith;waterileveld 232 feet above;theireactoriyessellf1angedthefeact'.sc coolantLtemperaturehis maintained;significantlyxbelow boiling _with:thettypicalitimeitolboil:significantlyiin excesslofitheToneihour; Continued 1decayiheat removal assuresithatitbefeactot; rem & inszi n; a i safelyicool ed condition,7andfadequate boron mixing assures;thatlan inadvertent criticality,would;not;occuG 8 04 A This change eliminates the option of securing RHR prior to initial criticality since initial criticality has - m already oM.noteKnoavnymMicable 0 3,9 13 1 mantz perations:and;1siobsoletejcycle; o dependent information;d Thi.s typetofeinformation may;be removed underlan1 administrative 1(A): change 1since no, technical changes'(either actual:oriinterpretational)Lare;being made! l - CPSFS Description of Changes to CTS 3N.9 6 72 9/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-17 APPLICABall'iY: CP, DC REQUEST: CTS 3/4.9.9, Control Rods DOC 10-01-R According to the DOC, the CTS requirements in 3/4.9.7 would be entirely relocated to an unspecified licensee controlled document. In addition, the DOC does not address adequate justification as to why the relocation is acceptable. Comment: Provide additicnal justification as to why this relocation is acceptable and identify the name of the licensee controlled document containing this requirement. This requirement shall be relocated to a licensee cont.*olled document controlled by 10 CFR 50.59. FLOG Response: DOC 10-01-R has been revised and a Technical Specification Screening Form has been prepared to provide justification as to why the relocation is acceptable. This justification shows the LCO provides assurance that adequate water will be present for iodine removal in the event of a FHA. Control rod movement is not associated with FHA and this LCO does not address inadvertent criticality. This LCO is not associated with any instrumentation used to detect significant degradation of the reactor coolant pressure boundary. It is not associated with any variable design feature or restriction which is an initial condition of DBA or transient analysis. It is not a pcrt of the primary success path for mitigation of a DBA or a transient. The format for specifying the location of relocated requirements (in Enclosure 3B of the conversion submittal) was found to be acceptable by the NRC technical specifications branch j reviewers during telephone calls on June 25 and June 30,1998 The document indicated in Enclosure 3B is controlled by inCFR50.59. ATTACHED PAGES: Enci 3A 6a & 6b i Att 21 47 & 48 l l l l l l I j

CHANGE NUMBER HS]E DESCRIPTION 8 05 Not used. 8 06 M This change adds an additional surveillance requirement to verify correct breaker alignment and indicated power ava'lable at least once per 7 days. This change adds a more stringent TS requirement which is appropriate and consistent with NUREG-1431 Rev. 1. 9 01 A Not applicable to CPSES. See conversion comparison table (enclosure 38). 9 02 LS-7 Not applicable to CPSES. See conversion comparison table (enclosure 3B). 9 03 LS 8 Not applicable to CPSES. See conversion comparison table (enclosure 38). 9 04 LS 15 Not applicable to CPSES. See conversion comparison table (enclosure 38). 10 01 R This change relocates the current TS requirements concerning reactor vessel water level for movement of [ control rods] to a licensee controlled document. This; specification pl aces:a g oWorilimit;onM.he 0 3*9 17 amountiof2WaterJaboveltheitoplofithelfuel assemblies'in L thelreactotivesselfducing; movement pficontrolirodsgTheiBasesZstatelthetithisiensuresj.the WateKremoves199; percent [ofithelassumedi10; percent;1odihe gapiactivity;releasedifton;thelrupture3fJaniirradiated fuel:assemblyLinithelevent:ofra:fue11 hand 11ng: accident (FHA)Iduring;corelalterations;IHowevetRtheisovement:of contrplirods;is;notiassociatedMth;thelinitia11 conditions ofian1FHATfandithe18ases;dolnotladdressiany; concerns regarding !!nadvettentIctiticality which;could ;1eadito ia  ! breach:ofithelfuelitod claddingg2Inadvertenticrjticality during mode:611sipreventedi by; maintaining: proper; boron concentratjonljn;thelcoolantlin accordance;With[LC0:319;11 1 Trisiptoposed iTSirevj sion1relocatesicoquirementszwhich[do j pos:meetLthelTSicrjterja;1nJ10CFR50136(c)(2)(11);1to ) documents;withle.stpbl ished; control t prograns nThi s regulationiaddresses;the:scoperand:purposeiof3S min doingisoEjtisets] forth"aj specificisetioflob, ject 19 eritertagoridetermining:whichrif atoryyequirements:and operating:restrjetionsishocidfbenneluded;jn;the;TS7 Relocation:of1theseirequirpmentsiallowsithelTSito: be reservedIonly!fpt;those;,conditionsJot211mitationsrupon reactor; operation M ich areinocessary,tolobviate'the possibility;oftanlahnormalisituation:orgevent giving: rise j to;aniimmediateithreatito;the;publichealth;and; safety CPSES Descr&nton of Changes to CTS 3N.9 6a 70988 _ o

CHANGE WMBER HSBC DESCRIPTION thereby; focusing; thel scopeiof;the;TSE An 'eyaluation;of g,3,9,17 j theiapplicability;oflthese criterialto;thjs i specifjcationtjs1provided;jn

Attachment:

21; To:ensurelanlapproprjate;1eveUoficontro1 E these i requirements y11E beIrelocated ito:1):documentsithat;are subjectiteltW provisjes;oflRCFRM 59 E2)fother licenseeEdocumentsjwhichthavelsis11arJtegulatory; controls (e;g;%theAualjty:Assurancellan;1as.: described;in:the ESARZ Which iisicontro11ed; by;10CFR50.54a)'2orj 3)lto prograssithatlere;contro11.edivjalthe; Administrative Control sisection:otthelieprovedlTS TGThe;1denti fication pfithe!specificI11censee:contro11eddocument;containing thisirequirementti.s!provided jnLEnclosurel38;ofithe conversionisubmittala comp 11ance31thithe;ralocated Requirements; willinot;be affected;bymthis;;;ptoposed: change:tolthelcurrent[ Technical Specificatims.nTheyequired; periodic:sur9111ancesy, win conitaueltolbe;performedJtolensCNEthatl11mitson parameters?arejsajntained.Z;JherefomErelocation;otthese requjramentsiw1111have]ios. impact l:onisystenloperab111tylor the; maintenance of[contro11 edj patanetersiwithinil imitsi l CPSES Description of Changes to CTS 3N.9 6b 7/29/98

TECHNICAL SPECIFICATION SCREENIN3 FORM (1) TECHNICAL SPECIFICATION 3.9.9.2 WATER LEVEL - REACTOR VESSEj, (CONTROL RODS) Applicable MODES: During movement of cortrol rods within the reactor vessel while in MODE 6. (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to: YES NO _ 1 (1) Installed instrumentation that is used to detect, and indicate in the control rooms significant abnormal degradation of the reactor coolant pressure boundary. _ .1 (2) A process variable, design feature, or cperating restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (3) A structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. _ 1 (4) An SEC which operating experience or probabilistic safety assessment (PSA) has shown to

• significant to public health and safety.

If the answer to any one of the above questions is "YES", then the TS shall be retained in the TS. If the answer to all tour of the above questions is "NO", the TS may be relocated to a controhed document. (3) DISCUSSION This specification places a lower limit on the amr ? of water above the top of the f :91 assemblies in the reactor vessel during movement (,f control rods. The Bases state that this ensures the water removes 99 percent of the assumed 10 percent lodine gap activity released from the rupture of an irradiated fuel assembly in the event of a fuel handling accident (FHA) during core alterations. However, the movement of control rods is not associated with the initial conditions of an FHA, and the Bases do not address any concems regarding inadvertent criticality which could lead to a breach of the fuel rod cladding. Inadvertent criticality during MODE 6 is prevented by maintaining proper boron concentration in the coolant in accordance with LCO 3.9.1. The TS requirements for water level - reactor vessel are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPD; l therefore, this TS does not satisfy criterion 1. l l The water level - reactor vessel (control rods) TS are not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, these requirements do not satisfy criterion l 2. Attachment 21 47 t-_ _ ________ __-.

i i Th3 TS for watrr Irv:1 - rsector vassil do not apply to an SSC that is part of the l primary success path and which functions or actuates to mitigate a DBA or transient i that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, these requirements do not satisfy criterion 3. The reactor water level during movement of control rods while in MODE 6 has not been shown to be significant to public health and safety by either operating experience or PRA. The reactor water level during movement of control rods while in MODE 6 is not modeled in the CPSES IPE. However, there is no indication that this function would be identified as risk significant if it was modeled in PRA models. Therefore, this TS does not satisfy criterion 4. (4) CONCLUSION This Technical Specification is retained.

                     .X.      The Technical Specification may be relocated to a licensee controlled document.

9 Attachment 21 48 L_____-___--_-_______--__- _ _ - - _

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O3.9-18 APPLICABILITY: CP, DC REQUEST: CTS 3.9.9.1 Applicability DOC 10-03-LS18 CTS requires movement of unirradiated fuel when there is irradiated fuel in the core. The licensee proposes to revise the applicability such that it applies only when irradiated fuel is moved. There is not any technical discussion provided in the DOC to justify this change. Comment: Please provide technical justification in the DOC as to why this is technically acceptable and how it applies to current licensing 5csis. FLOG Response: The text of DOC 10-03-LS-18 is avised to add the following statement:

                     "The purpose of [ CTS 3.9.9.1) is to assure that sufficient water is present to remove 99% of the release of 10% iodine gap activity during a FHA. A FHA wuld result during movement of an irradiated assembly due to the drop of the assembly on the floor of the refueling cavity or the drop of any load on the core. A fuel assembly and its handling tool are the analyzed limiting load for the FHA. However, dropping a new (unirradiated) assernbly on the cavity floor would not result in an FHA and dropping a new assembly on the core is no different then dropping any other load on the core. This condition is already addressed by the programs associated with NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants," and is redundant to this LCO requirement. The FHA resulting from the drop of a new assembly would occur at the active          l irradiated fuel already loaded in the core and maintenance of 23 feet of water above the flanne (this LCO) would be unnecessarily conservative."                                                  j l

l ATTACHED PAGES: Encl 3A 7 I I l l l I L-_________________________-._.------_____ _ _ _ _ _ _ _

                                                                                                                                                                                       )

CHANGE EMBER NSHC DESCRIPTION 10 02 LS-22 This change deletes the surveillance requirement to verify water level within 2 hours prior to the start of movement of [ ] fuel assemblies. This is acceptable because the LC0 must be met at the time that movement of [ ] fuel assemblies is performed. 10 03 LS-18 Revises applicability such that it applies only when irradiated fuel is moved. The current TS also applies ' to movement of un irradiated fuel when theca 4 g irradiated fuel in the core A ne purposetof}[ CTS 0 3.9 18

                                  L aJasgo.===.dretthetisufficient1waterfis present tol remove;995;ofithe;releassic f;10Liodine^ gap activity;duringla:FMZKFHA.lcouldiresultidurIng movement ofianijrradiated assembly,duettoltheldroplof theiassembly                                                                      t on;thelfloorlof;thelrefueling;cayjtylordthe: drop of many load;onitheJcore3 Mfueli assembly, andlits'..handl ing; tool areithelanalyzed d imiting ;1oadifor,;theJHAEHoweveri dropping stnew;(unirradiatedEassembly'on;the; cavity l floor wou1d not;resultiin;an FHA andLdrppping;aLnew; assembly;on thelcoreds"no;different;thenf dropping;anylothetiloadlon thelcoredThe FHAiresulting;fromithe;droplofialnew assemblyLwouldiocentat3theiactivelirrudiated;fuellaiready loadedin;thecore;and;maintenancecofi23;feetiof; water pbove;thelflangeg(this:LCO)twouldibe' unneces<acW -

conservat;1ve! 10104 A Not f appl icabl.elto;CPSESMSee: conversion : compari son ; table (enclosure 138)i 11 01 LG This chag; ;;;;difies th; ; applicability t; during 0-3.9 21

                                 ;;;;;;; cat of irr;diated fuci ess; blics in the fuel stcrage [;rees] to k ;;nsist;nt with the fuel
                                 ";ndling Accident (r"M . The portions of this r;quir;;cnt                                                                                              l eppHesble-t; wherever irr;disted fuel is in fuci ster ge                                                                                               l

[er;;s] will bc ;r.;;;d t; ; licens;; centrolled docu cat.  ! Thi;, chage is ;;nsist;nt with N" REC 1431, n;v.1 ;nd 1 r;;;;vcs details that era not r; quired t; be in the TS to  ; prctcct the health and s;fety of the public while J r;teining the b; sic li;;;iting ;;nditions for ep;retien.;The Appl icabil ity;i sirevised;toyequireltheWaterd eve 11be l maintainedionly when3oving;jrradiatedifueWsThe bounding design;basfs;fue11 handling accidentijnithefuel; storage [ area]M assumeslandirradiated ;fuellassemblyii s1 dropped ontolan] array;ofd rradj ated; fueKassembliesJ seatedlinl the L fueUstoragedrack C EThe revised:Applicabilityljs consistent;withithis;designbasjszaccidenti(DBA)';dThe waterilevellrequiremennis;necessary;to; mitigate;the consequence of;th1G OBA E Novingithel requirement 1to maintain!fueUstoragej[ area) waterilevel;whenever irradlatedifuellassemblies arelin;the;fuelsstorage [ area) Provides;fonLeonservati ve; pl ant! operations J Moying ithi s l CPSES Descriptwn of Changs to CTS 3N.9 7 7/29/98

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q3.9-21 APPLICABILITY: CA, CP, DC, WC REQUEST: CTS 3.9.10 Applicability and 4.9.10 (Comanche Peak) CTS 3.9.11 and 4.9.11 (Callaway)

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CTS 3.9.11 Applicability and 4.9.11 (Diablo Canyon) l CTS 3.9.11 and 4.9.11 (Wolf Creek) DOC 11-01-LG { i The CTS requirement, applicable to whenever irradiated fuels are in the fuel storage racks, is proposed to be relocated to an unspecified licensee controlled document. The DOC does not provide any technical justification related to this reloco5n. Comment: Revise the DOC by including the justification for the relocation and identify the licensee controlled document containing this requirement. This requirement shall be relocated to a licensee controlled document controlled by 10 CFR 50.59. FLOG Response: Identification of the licensee controlled document containing this requirement is identified in Encia ~re 3B of the conversion application. This was discussed with the NRC technical specification oranch reviewers on June 25,1998 and determined that the information provided in Enclosure 3B was acceptable. DOC 11-01-LG has been modified to include the following information: "The Applicability is revised to require the water level be maintained only when moving irradiated fuel. The bounding design basis fuel handling accident in the fuel storage (area] assumes an irradiated fuel assembly is dropped onto an array of irradiated fuel assemblies seated in the fuel storage racks. The revised Applicability is consistent with this design basis accident (DBA). The water level requirement is necessary to mitigate the consequences of this DBA. Moving the requirement to maintain fuel storage [ area] water level whenever irradiated fuel assemblies are in the fuel storage (area] provides for conservative plant operations. Moving this information maintains consistency with NUREG-1431. The information is moved to a licensee controlled document which is controlled by a 10 CFR 50.59 changa process." ATTACHED PAGES: Encl.3A 7 & 7a l N________________--_______

CHANGE NUMBER NSHC DESCRIPTION 10 02 LS-22 This change deletes the surveillance requirement to verify water level within 2 hours prior to the start of movement of [ ] fuel assemblies. This is acceptable because the LC0 must be met at the time that movement of [ ] fuel assemblies is performed. 10 03 LS-18 Revises applicability such that it applies only when irradiated fuel is moved. The current TS also applies to movement of un irradiated fuel when there is irradiated fuel in the core. The;purposelofi[ CTS 0 3.9 18 319.9il]Ms;tojassuref thatisufficientswaterlisipresent toiremove1995;of;thejreleaselofM0*Ljodine gap activityiduring;alFtlAEAJHA;couldiresultfldurjng; movement otiandrradiated; assembly;.dueltolthe;droplof3thelassemb]y on;thGif]ootof;thelrefuelingicayjtyloritholdroploffany J oadionithe[coreM Alfuell'assemblyIandM ts1 handling 1 tool are 3he"analyzedilimitingD oadifotithe M B E ttowevet? dropping:alnew;(unirradiated);assemblyfonitheicayity;f]oo[;

                                      @ul.dinotiresultUn:anIrlA2and droppingtainew! assembly;cq the; core 11 s(noldj fferentithen: droppingIanyiotheQload;on the:coreETheWHA?resultingifras;the;dr.oploffafpow assemblyLwould;,occurCat;theiactive;1rraff atedlfuellalreatty loaded iipithe ~ core [andl maintenanceiofj 2Rfeettof; water, above? thei f1 angel'L(thi s! LCO)f wouldjbe; unnecessarily conservative!

10294 8 No t1 appl icable: to1CPSESM See:conversionlcomparj son; table (enclosure?38)!. 11 04 LG This charge ;;difies th: ;pplic;bility t; during 0-3.9 21

                                      ;;;;.nnt of irredicted fuel essa..Liiss in the fuci storag: [;rces] t; be censistcat with the fuel
                                      !!;ndling Accidcat (TitM. The pertions of this rcquir;n nt}}