Text

Forwards Addl Info or Supporting Documentation Not Provided in Original RAI Response,Answers follow-up Questions or Addl Changes Identified by Licensee Re 970515 Request for Amend to License NPF-42
	ML20198H105
Person / Time
Site:	Wolf Creek
Issue date:	12/21/1998
From:	Muench R; WOLF CREEK NUCLEAR OPERATING CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	ET-98-0107, ET-98-107, NUDOCS 9812290239
	Download: ML20198H105 (155)
	v • d • e

{{#Wiki_filter:. m-, (. I 9 W$4.F CREEK NUCLEAR OPERATING CORPORATION [. Richard A. Muench Vim President Engineenng DEC 211998 l ET 98-0107 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station Pl-137 Washington,' D. C. 20555

Reference:

1) Letter WO 98-0078 dated August 5,

1998, from C. C. Warren, to-USNRC

2) Letter ET 98-0071 dated August 28, 1998, from R.

A. Muench, to USNRC

3) Letter ET 98-0078 dated September 24, 1990, from R. A.

Muench, to USNRC 4, Letter ET 98-0078 dated October 16, 1998, from R. A. Muench, to USNRC

5) Letter ET 08-0098 dated December 2,

1998, from R. A. Muench, to'CSNRC

6) Letter ET 97-0050, dated May 15, 1997, from R.

A. Muanch to USNRC

Subject:

Docket No. 50-482: Follow-up Items Related to the Proposed Conversion to the Improved Technical Specifications Section f 1.0, 3.1, 3.2, 3.3, 3.4, 3.7, 3.9 and 5.0 i ' Gentlemen: Wolf Creek Nuclear Operating Corporation (WCNOC) requested an amendment to the Wolf Creek' Generating Station (WCGS) Unit 1 facility operating license -(NPF-

42) by incorporating changes to the WCGS Technical Specifications (TS) as provided in Reference 6.

The ' NRC staff requested additional information regarding Section 3.1, "Reactivtty Control Systems," Section 3.2, " Power Distribution Systems," Section 3.5, " Emergency Core Cooling Systems," Section

3. 9,

" Refueling Operations," and 4.0, " Design Features," which was provided in Reference 1. The NRC staff requested additional information regarding Section

1. 0, "Use and - Application," Section

2. 0,

" Safety Limits," and Section 3.0, " Limiting Conditions for Operation Applicability / Surveillance Requirement j Applicability" which was provided in Reference 2. The NRC staff requested additional information regarding Section 3.4, " Reactor Coolant System," and / Siction 5.0, " Administrative Controls," which was provided in Reference 3. The~ NRC staff requested additional information regarding Section 3.7, " Plant j. Systems," which was provided in Reference 4. In addition, the NRC staff requested additional information regarding Section 3.3, " Instrumentation," D/ which was provided in Reference 5. The Attachments to this letter provide / (1) additional information or supporting documentation not provided in the i original Request for Addition Information response, (?) answers to follow-up questions, or (3) _ additional changes identified by the icensee. This letter and Enclosure are not a supplement to Reference 6 and have not been reviewed by the Plant Safety Review Committee or Nuclear Safety Review Committee. A supplement to Reference 6 will be provided at a later date. 9812290239 981221 PDR ADOCK 05000482, P PDR 't' P.O. Bo$411/ Burkngton, KS 66839 / Phone: (316) 364-8831 ao An Equal Opportunity Ernployer MF/HC/ VET I

i l l l -ET 98 0107 Page 2 If you have any questions concerning this response, please contact me at (316) 364-4034, or Mr. Michael J. Angus, 316-354-4077.

o, Very tr' y yours,

[ ' Rich rd A. Muench l ' Attachments -Enclosure-RAM /rlr cc: W. D. Johnson (NRC), w/a E. W. Merschoff'(NRC), w/a Senior Resident Inspector (NRC), w/a K. M. Thomr_s (LAC), w/a 1 i 1 4 t {

. ~ i STATE OF KANSAS ) ) SS COUNTY OF COFFEY ) i Richard A. Muench, of lawful age, being first duly sworn upon. oath says that j he is Vice President Engineering f Wolf Creek ruclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief, i l By Richard I. Muench Vice President Engineering SUBSCRIBED and sworn to before me this c1}U day of Decembe, 1998. hfM Notary Pu 1'c O caenNEWWWWI g m 7/r/soox Expiration Date ~~T~u \\q ( doo 2 i l i o

AttachmInt 1 to ET 98-0107 Page 1 of 1 The item numbers are formatted as follows: [ Source] [lTS Section]-[nnn) Source = Q - NRC Question NR - NRC Follow-up Question CA - AmerenUE DC-PG&E. WC - WCNOC CP - TU Electric TR - Traveler ITS ltem Number Applicability Enclosed 1.0 Q 1.4-1 TSTF-267 TSTF-270 CA, CP, DC, WC YES 3.1 Q 3.1-25, WOG-105 CA, CP, DC, WC YES 3.2 Q 3.2-6, TSTF-241 CA, CP, DC, WC YES 3.3 TR 3.3-007 (NEW), TSTF-311 CA, CP, DC, WC YES 3.3 CP 3.3-011 (NEW) CP NA 3.3 WC 3.3-019 (NEW) WC, CP, DC, CA YES 3.4 Q 3.4.1-2, TSTF-282 CA,WC YES 3.4 Q 3.4.5-2 CA, WC YES 3.4 Q 3.4.11-3, TSTF-113 CA, CP, DC, WC YES 3.4 - Q 3.4.12-1, TSTF-280 CA, CP, DC, WC YES 3.4 WC 3.4-010 ,CA NA 3.4-WC 3.4-011 (NEW) WC YES 3.6 Q 3.6.3-23, WOG-126 CP,DC NA 3.7 O 3.7.G-1 CA NA _ _3.7 Q 3.7.1-4, TSTF-235 CA, CP, WC YES 3.7 Q 3.7.10-14, TSTF-287 CA, CP, WC YES 3.7 Q 3.7.13.2-2, TSTF-287 CA,WC YES 3.7 CA 3.7-ED CA NA 3.9 Q 3.9-7, TSTF-312 CA, CP, DC, WC YES 5.0 0 5.5-4, TSTF-308 ~ CA, CP, DC, WC YES

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 1.4-1 APPLICABILITY: CA, CP, DC, WC REQUEST: DOC 1-26-A ITS Example 1,4-4 ITS Example 1.4-5 JFD 1.1-3 JFD 1.1-11 Additional examples, Example 1.4-4 and 1.4-5, are proposed to be included in ITS. The DOC and the JFDs state that these ITS changes are to incorporate travelers WOG-74 and WOG-90. Comment: Provide the current status of WOG-74 and WOG-90. If WOG-74 and WOG-90 are not approved by the TSTF, then these changes should be withdrawn from the conversion submittal at the time of the TSTF rejectinn. If WOG-74 and WOG-90 have not been acted upon by TSTF, or have been approve: my the TSTF, but not approved by the NRC at the time the draft safety evaluation is prepaied, then these changes should be withdrawn from the conversion submittal. These changes will not be reviewed on a plant-specific basis. FLOG RESPONSE: (original)WOG-74 and WOG-90 have been approved by the TSTF and are designated as TSTF-270 and TSTF-267, resp?ctively. Both of these travelers have been submitted to the NRC and are under review. The proposed wording in TSTF-270 was modified from WOG-74, Rev. 2, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by these travelers. FLOG RESPONSE: (supplement) Based on the present status to the generic change process for the STS, it appears that travelers TSTF-260 and TSTF-270 will not be approved by the NRC in time to support the initial license amendments for the FLOG plants. In order to facilitate the issuance of these initiallicense amendments, an alternate approach has been developed which relies on the CTS, plant specific information and/or the NUREG but does not rely upon the travelers. This alternate approach it hereby provided as an interim submittal to allow issuance of the initiallicense amendments. The changes which rely upon the travelers can be processed in subsequent license amendments following approval of the traveler by the NRC. The FLOG believes that these travelers reflect correct rules of usage for the ITS and willinclude them in operator training and plant procedures pending approval of the traveler. At this time, Examples 1.4-4 and 1.4-5 of ITS Section 1.4 have been deleted and JFDs 1.1-3 and 1.1-11 have been revised to indicate "Not used." ATTACHED PAGES: Attachment No. 4, CTS.10 - ITS 1.0 Encl. 3A 5, 6 Encl. 5A Traveler Status page, 1.4-5,1.4-6 Encl. 6A 1,2,3 Encl. 6B 1, 2

-s. CHANGE NLABER hl2iG DESCRIPTION would be tested each month during any given quarter. 1 24 A The current TS definition of Site Boundary and Unrestricted Area are deleted to be consistent with NUREG-1431. These definitions are deleted on the basis that they are defined in 10 CFR 20.1003. 1 25 jdK2 A Table 1.2 of the current TS would become Table 1.1-1 in .the improved TS. The following changes would be made to {(t.t-9] conform to NUREG 1431. In new table 1.11, the notation "NA" would replace "0" under

Rated Thermal Power for Modes 3. 4, 5 and 6. This is a nontechnical change since with K,,, less than 0.99, thermal power would be zero anyway. For Mode 6, the temperature has been replaced with NA since there is no safety analysis basis for the value of 140*F specified in the current TS. Also for Mode 6, the reactivity condition has been designated NA since the value of 0.95 is specified in the Bases for improved TS 3.9.1.

The temperatures for Modes 1 and 2 are designated as.NA on the basis that temperature for these modes is dictated by the minimum temperature for criticality and the operating program for reactor coolant system Tava. A new note b has been added to Modes 4 and 5 GII' stating that thl@p6W)id) reactor vessel head closure 4,gj' bolts are fully tensioned, and a new note c replaces the _ note ap) lied to Mode 6. The new note c states that @ =~ g Ers@@e3 reactor vessel head closure bolts are less than m The new note c no longer specifies that fulTy tensioned 2 C.3 fuel is in the vessel because the condition of fuel in the I vessel ir addressed by the definition of the term Mode. This definition stipulates that fuel be in the vessel in N order to be in a " MODE." These changes are s 2 gq 1 26 A New sections 1.2.- 1.3, and 1.4 would be incorporated into the improved TS to be consistent with NUREG 1431. Section 1.2 provides specific examples of the use of the logical connectors Mil and QR and the numbering sequence associated with their use in the improved TS. Section 1.3 deals with the proper use and interpretation of completion times, and specific examples are given that will aid the user in understanding completion times. Section 1.4 deals with the proper use and interpretation of surveillance frequencies. Specific examples are given that will aid the user in understanding completion times. Specific examples are given that will aid the user in understanding surveillance frequencies as they will appear in the WCGS-Description of Changes to CTS 1.0 5 S/lS87

l i CHANGE j NUMBER HSHC DESCRIPTION improved TS. The proposed changes are administrative in i nature and by themselves are not technical changes. porgaptfig)thpgM. /h @##'d mp: -% iA.drrfzw] l 1 27 Not applicable to WCGS. See Conversion comparison Table j (enclosure 3B). ) i 4 1 \\ l 1 28 LG The current TS definition of CONTROLLED LEAKAGE is deleted to be consistent with NUREG 1431. Rev. 1. The RCP. seal water return flow limit is moved to a licensee controlled I document. @*.T 34 - 4 / o.a s. s -2.] l l 1 29 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). lot t-!.} } 1 30 A Consistent withG5549WA, the definitions of Channel Operational Test (C0T)o [ ] and Trip Actuating Device (pwe MmasN T[te.st,h,. Operational Test (TADOT) is expanded to include the 3 details of acceptable performance methodology. Performance of these tests in a series of sequential, overlapping, or total (G!DIBr@ steps provides the necessary assurance of appropriate operation of the@ channel. This change also makes the COT.+[ ] and TADOT definitions consistent with the current TS and the NUREG 1431 definition of channel calibration which already contains similar wording. @l% _ *

9 ".re*P a t M 1 31 N

o W d5Wnfosp o t,t_q r 1 32 A The definition of channel calibrations. COT. and TADOT is g i.g.2. I h% reworded to be consistent with TST@to clarify the phrase " entire channel" thus reducing the potential for inconsistent interpretation of the ohrase as experienced by a number of plants.fAsimitw c tmReam c3 peias.4 6e Ac.bhO {J-_ (t. e.pa. Ted., 5 * * ' ~ l 1 33 Thipcha r 'se he C 5 d in ion f re Alt ati ns delet or ni atio an c ser tiv ~ ns' te wi 1 wor a se in e fi tio we e du nt and eti the rds ces ot ter th mea ng f the de niti rh,-t wplicabe.a.to WC.GS. See. Corwerson Cempare im,TsM.a (.hcloswv 3B),- ( MotMc2W. b N5. 5ec Ce,nesicmComreSm ~ 34 tA 'i % TaLA. ( Endosunn.300, [.%:-c.tn 3.6 9 AT..sl 1-3S +hr u 1 - 41 NS.ERT 3A 3 WCGS-Description of Changes to CTS 1.0 6 S/15M l

( INDUSTRY TRAVELERS APPLICABLE TO SECTION 1.0 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-19, Rev. I hincorporated )fA % NRC approvedArof l Mi o-ed I.t-r2-g alepem 4 due. TSTF @ fFI Ikorated 1.1-9 @ krs.i-i1 l l (TSTfg Igrppritf fM Let.t-2.[ k-JV Ipeofppriftetf 1J49 {a. ] TSTF-111, Rev.h hNedporated 1.1-5 {at.. } I @,;&T M"gorated {h {G1.1-srl h, (R p f5Incorporated U [al.4-R MM Incorporated .1-11 \\Gn4-L I Tsrp-s2 inwporateL~ t.1-13 a 1 4 1 as.L.t-u \\ 1 %.o tnc,o9<ekeel c w p prORC m meds 1 l l l a 5/15/97

.. -.-.~. - -. Frequency 1.4 1.4 Freauency E S EXAMPLE'1:4:4 gg 4 g nud) SURVEILLANCE ~ REQUIREMENTS / SURVEILLANCE F QUENCY i l h15htbe On1 rformed in H00E 1 i Perform complete cycle of the valve. 7 days The interval continues, whether or t the unit operation is in MODE 1.~ 2:or 3 (the assumed Appl ability of'the associated LCO) between performances. As the Note modifies the r ired performance of the Surveillance, the Note is onstrued to be part of the "specified s Frequency " Should the, day interval ~betexceeded while operation is:nctin-1, this Note 4110ws entry into and operation;in H00ES'2 nd.3,to perform the Surveillance. T6 Surveillance is sti considered to be performed within the I "specified Fr y" if completed prior to entering MODE 1, Therefore, if't. Surveillance were not performed within the 7 day (plus the e ension allowed by SR 3.0.2) interval, but operation was t in MODE 1. it would not constitute a failure of the SR ortfa ure to meet the LCO. Also,. no violation of SR 3.0.4 occu' when changing MODES, even with the 7' day Frequency not met, rovided operation does not result in' entry into MODE 1. Once t unit reaches H00E 1, the requirement' for the Surv 11ance to be performed within its specified Frequency app es and would require that the Surveillance have been formed. If the Surveillance were not performed prior to H00E , there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would applyWapp@gAfad{vipntjeh oM. [q i,4_ l I 1 (continued) i WCGS.Sfark up ofNUREG-101 - ITS 1.0 1.4 5 S/15/97

l Frequency l 1.4 1.4 Freauency m l EXAMPLES EXAMPLE 1.4 5 psugy [Q i.4 I { gg (Continued).. SURVEILLANCE REQUIREMENTS / SURVEILLANCE FREQUENC Verify each containment isolation manual Prior t valve is closed. enter g H00E.4 fr E 5'if performed thin the previous 921dayQ j / In Example 1.4 5. the "specified Frequ cy" is measured from when the Surveillance was last performed.. hould the interval be exceeded. the SQrve111ance is not r uired to be performed until certain conditions are met. Theu rve111ance is allowed to be delayed until prior to entering E 4 from HODE 5 if the 92' day "specified Ft equency" has exp1 The 92 day interval'may be extended to 1.25' times the s ted interval as allowed by, SR~ 3.0.2 for operational"flexibilit. Therefor (, if the.Surve111ancerwers not performed within the 2 day (plus the extension allowed by:SR 3.0.2) interval, but o ration was not transitioning from MODE 5 to. MODE 4, it would t constitute a failure of the SR or a failure to meet t CO. The next time the unit proceeds from HODE 5 to MODE'4 the surveillance would be required to be performed prior o the transition. The measur nt of this interval continues at all times, even when the is not required to be met per SR 3.0.1 (such as when the nt is inoperable, a variable is outside specified limit, or the. unit is outside the Applicability of the LCO). If the onditions in the Frequency are met and the-interval s ified by SR'3.0.2 is exceeded without the Surveillance having een performed and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable. I I WCGS-Mark-up ofNUREG-1431 -ITS 1.0 1.4-6 5/l5/97

DIFFERENCES FROM NUREG 1431 Section 1.0 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431. Revision 1, to make them plant specific 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 6B, text in brackets "[]", indicates the information is plant specific and is not common to all the Joint Licensing Subcomittee (JLS) plants. Empty brackets inideate that other JLS plants may have plant specific information in that location. CHANGE NLM ER JUSTIFICATION 1.1-1 The NUR - 431 .1d initio f Chan Calibpwion s es,"TheD l lC CAL TION all en pass t entire annel nelud the uir

ensor, arm. i erlock,; splay, trip unctio T s
chan clarif what compass the en e chan by r rdin initio o state 'The CF EL CALI TION.

11 en pass se compo s, such s sensor, alarms display, and t p fun Cha[ns,/ 1o r red to > form the pecifie safety f ction ." T n ration est, a rip Ac ting Dev'ce Oper tiona est d6finit(ons are si arly revi This change i consist nt wi TSTF 64. tusEArp-Ib 1.1 2 Not used. nly uir to M00 c ~ o lo iA-i i wi trayeer 1 7 1.1 4 Not used. 1.1-5 The defin_itions for ESF Response Time and RTS Response Time G35ifBib lGL t-+\\ re_ vised)tos stitu the rd " ertf d" i lieu f "me 'ured" ~ c 451 en ith er 1reme s of URE 431 3.3.16 d SR .3 .1 Thi chan woul ens e co ste.y bet en t def iti R ponse ime d th requ emen to riodi ally erifv es s i is wi in l' its. his ang isfconsistent with traveler TSTF-m 111. Q 1.1 6 The definition of the Pressure and Temperature Limits Report (PTLR) would be revised to include the maximum allowable PORV lift settings and arming temperature associated with the [ Low Temperature Overpressure Protection (LTOP)] System, and to be consistent with the COLR definition. Improved Technical Specification 3.4.12 states that the PORV lift settings are specified in the PTLR. The current definition for PTLR does not identify these lift settings as being contained in the PTLR. WCGS-Differencesfrom NUREG-1431 - ITS 1.0 1 S/2S/97

sn aMA, % Pn.R sah;%*n i%hulu %. P6e\\1 CHANGE IG q3 %,h w% h.w.texscw9 NUMBER JUSTIFICATION 6 t rs.se.c. tin s.s..e,. g The [LTOP] arming temperature was added to the PTLR. since changes in the heatup/coolaown figures could change the arming temperature. This change corrects the PTLR definition to be consistent with all of the requirements contained in the PTLR. Referenced methodologies for the PTLR would contain the methodology used to develop the heatup and cooldown figures, as well as the methodology for deve]oping the [LTOP]; setpoints. This change is consistent with traveler GQif'IrPAps-D.V @ st e_-2. g IQu-S] 1.1 7 Not applicable to WCGS. See Conversion Comparison fable (Enclosure 6B). -yon. _q } (gThe re or vess ead c ure requir ts for ES 4 a 6 1.1-8 are arified .e pr sed c ge revi footno b for ES and i o refer o " Req ed rea r vessel ead clo re bolt full tensio " and e c fo E6i evised read " uire reactor l ' ve head osure b s less t n fully nsione The ransiti int be n MOD and 6 w d also clarif as curring n/ the r uired r tor vesse ead clo e bolt re le than f y j t ioned. e requir number of losure t <., ch may less an the otal n , is estab shed by alysi that d nstr s adequ e 0 ring ression prevent eakag nd ens st 1 S on III sJ ss limits r affec c nts are not exc

88. rWepucMA.b ucces.s)es.

1 T is change 4 s consist.e with T (c nvasim compm.iacm-rae cero.sure. 4.s)g 1.1 9 Consistent with TSTM. the definition of Channel Operational l

5 Test (C0T),,[ ] and Trip Actuating Device Operational Test (TAD 0T) is

% q.9*

espanded to include the details of acceptable performance methodology. Performancs of this test in a series of sequential, overlapping, or Y total @ steps provides the necessary assurance of appropriate operation of the@channelf This change also makes the COT definition consistent with the definition of channel calibration which already contains similar wording. fick.rdevi9 J 1.1 10 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). lQ I 4-l k 1.1 11 Add /n xam e to SS ion 1 # to c rif surve ance reque ies] ,tKat ec ing on th a s cifie fr ency a d pla condi ions.

Th TS ntai many urveil nce F quen es th are nting t on i

g gg th a speci ied Fr,quency" and p nt co dition. For examp }"Wi n7 ys pri r to t initi tion Phys s Tes s," a "Pri to e ering E4 om 5 if ot pe ormed ithin he pr vious 2 I ) ays." hese equen es do ot fa clear y und any the xi g ec 'on 1.4 ample. The opos exam e is n ded mak cle f WCGS-Differencesfrom NUREG-1431 - 1TS 1.0 2 S/lS/97

1 CHANGE i NUMBER JUSTIFICATION ~- that SR .0.2 ex stion ' 1.25 mes he s cifi frequ ncy a i to spec ed Freq cy, 2) t int val ai d to l i pe orm missed rveilla e by 3.0. ppl s. i i i S .0.2 i clear th the. 5 ext sion ay be pplied.o "t nterv specifie in th reque y." the posed ange oes t fchan the in t of Spec cati s. .0.2 lie fa j i I 5 veilla is n perfo wit the 'specifi Fr ency Agai l l the ex ed not c ge t ntent f the S cif atio but o /mak clear app ation SR 3 .2 and .3 Sur illan with F uenc s tied plant onditi s. Thi cha wil elimi [e l / conf on and sappli ion of he ITS w ens e co st \\a 1 cation f SR 3. 2 and.. 3 to t se so Surv ' 1 ce ( requenc s. This hangeilconsis nt w'_h_tr ler o L4-1 \\ h2. tussa:r GA-3gj ng g,o,og q ta se a r d - * > L a 2.6.1 - a 1 l i l l l i l WCGS-Differencesfrom NUREG-1431 - ITS 1.0 3 S/15/97 i

l o i l INSERT 6A-3a TR 1.0-006 1.1-12 The definition of CHANNEL CAllBRATION is revised per TSTF-19 to move details of RTD and thermocouple calibration to the ITS 3.3 Bases associated with calibration of the components. 1 INSERT 6A-3b 0 3.6.1-6 l42.u-L\\ 1.1-13 Traveler TSTF-52 @ g B 6s dn.# 0 deletes the definition of La. l Since La is defined in 10 CFR 50 Appendix J and ITS Section 5.5-16, Containment Leakage Rate Testing Program, it is redundant to include La as a definition. As described in NUMARC 93-03, " Writer's Guide for the Restructured Technical Specifications," Specification 1.1 is a list of defined terms and corresponding I definitions used throughout the Technical Specifications. La is not used throughout the Technical Specifications and is t lined in l Section 5.5-16. l f l l I' L l. I.

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page I of 2 SECTION 1.0 DIFFERENCE FROM NUREG-1431 REV.1 APPLICABILITY I NLM ER ~ CRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 1.1-1 ' Tid,s c Id clarif at enc sses t entiM Yes Yes Yes Yes ) cha 1b ng defin on to st

e. "T

,,,3 g .W.I TI shall e ass t ec nts I p in c 1 uch as sors. a . di ays. ip ti . that e requir to perf the r s ift afety f tion (s) The CGT and T (definit ons are imilarly evised._ __ l 1.1-2 Not used N/A N/A N/A N/A 1.1-3 ex f Wir N/A Ms MrA ydr N/A MS M/A Not itsed, Q l.4 1 l 1.1-4 Not used N/A N/A N/A N/A i 1.14 The defini_tions for ESF Response Time and RTS Response Yes Yes Yes Yes Ig Time W evised.fo s titu tt wor o 43 3.3 .1 nd '3.3 .1Q j 1.1-6 The definition of the Pressure and Temperature Limits Yes Yes Yes Yes Report would be revised to include the maximum allowable PORV lift settings and the arming temperature associated with the system. and to be consistent with the COLR definition. (Mbp clavi (U/o ha. defiv6hoA of CHA&3El cat.tBRATtN, cr$T, amA rADcrT b3 <=placq 1ha a, bigu.ou.s. g with "devt$ts tJ tha. chewd <egpd -ft:wa cha+srui. OPEJRAe* L.tT'f. dh* AchN Uxyc. Test de-fwAhab cs dmilaib vevisca. S/158 7 WCGS-Conversion Comparision Table-ITS 1.0 r

' CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 2 of 2 SECTION 1.0 DIFFERENCES FRON BRIREG 1431 REY. 1 APPLICABILITY NUlBER DESCRIPTION DIABLO CANYON COMANCE PEAK. n0LF CREEK CALIAiAY - Q t> & - = p ]. xUWD 1.1-7 The definition of Channel Functional Test in the Yes Ido - Idot part 100 - Not part. 'No.- 100t part current TS will be retained in the taproved TS. This of current TS. of current TS. of current i TS. v definition is not in IRREG-1431 Rev 1. 1.1 8 Note b is revised to refer to the "Itequired reactor h>5o .Ygfe do Ws 09 Yes(Gt.t-9 l t vessel head closure bolts fully tensioned" and note c is revised to read " Required reactor vessel head closure bolts less than fully tensioned." L Yes_ __ Yes Yes Yes y The definition of Channel Operational Test (C0T).i[ ] Fmaelwr Q %=t gq g,,,g g [ 1-1-9 r and TA00T are expanded to include the details of acceptable performance methodology. Performance of gnave.va.13aut, + this test in a series of sequential. overlapping. or total M teps provides the necessary ass Ar nce _2 ~ eti@] u g eju.a.,suspi of appropriate operation of the %jhanney rela 3 c t 1.1-10 This change is based on the current TS definition of 100 - Ilot part No - Not part No - Yes i CONTROLLED LEAKAGE. This change is a clarification of current TS. of current TS. Maintaining l ISTS wording. only and does not affect the way RCS water inventory balances are performed. 1.1-11 to ar ,Vgr N/A We /A Yqlb v/4 yde,.3 1 y pl iti s. Mot (1seL [ IQ l.4-1 L i tusEKT 6B-h [ ~TW.1 o -oo G j l i ( l.1 -_l3 tOsE5trGB ' htl q.s.c..t c } 1 l MSM7 i WCGS-Conversion Congparision Tame - 1TS 1.0 l

INSERT 6B-Za TR 1.0-006 . TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 1.1-12 The definition of CHANNEL CALIBRATION is Yes Yes Yes Yes ~ revised per TSTF-19 to move details of RTD and thermocouple calibration to the ITS 3.3 Bases associated with calibration of the-components. INSERT 6B-2b 0 3.6.1-6 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 1.1-13 Traveler TSTF-52CAfr>fCJeif"gA deletes Yes Yes Yes Yes the definition of La. M nce La is defined fq u.i-(., \\ in 10 CFR 50, Appendix J and ITS Section 5.5-16, Containment Leakage Rate Testing Program, it is redundant to include La as a definition. -.,--__a__-_._---___._-__----_.__--..-._-_-_...___--__.-___.--.-..._----._a___.--

l l ADDITIONAL INFORMATION COVER SHEET 1 I ADDITIONAL INFORMATION NO: Q 3.1-25 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 3.1.4 Rod Group Alignment Limits (All FLOG Plants) 1 JFD 3.1-16 l l Comment: Inclusion of SR 3.2.1.2 to Required Action B.2.4 is approved; ensure OG l submit WOG-105 as a TSTF change request. 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, "Fo (Z) (F Methodology)." However, y this additional work has no impact on the manner in which the FLOG has l incorporated this trave'er's additional restriction. The TSTF will be submitted to NRC expeditiously. FLOG RESPONSE: (supplement) WOG-105 has not become an NRC approved traveler in I time to be incorporated into the license amendments for the FLOG. In lieu of incorporating this traveler, the STS markup and ITS is being justified based on the CTS. In the CTS, the associated action requires that the Heat Flux Hot Channel Factor be verified without specifying whether the steady state value (Fo ) or the transient value Fo* or both are the correct ) C values to be verified. The CTS has been understood to require the verification of both values and the ITS is hereby revised to match this understanding of the CTS. ATTACHED PAGES: Attachment No. 7, CTS 3/4.1-ITS 3.1 l Encl. 5A Traveler Status page Encl.6A 3 Encl. 68 3 Attachment No. 8, CTS 3/4.2 - ITS 3.2 l Encl. 5A Traveler Status page j Encl. 6A 3 4 i i l l

t INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated ' 3.1-15 NRC approved. ITS Special Test Exceptions 3.1.10-is retained and j renumbered as 3.1.8, t consistent with this l traveler and TSTF-136. I' TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. f ne s.ser l l TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. (TSTF-107,4t. D Incorporated 3.1-6 [4 2.1 -s s-l _ __~ N:: E::: p s:;d -NA-- ~ Net NRC approdd nW l' TSTF-iO8, Rev 1 lacorporyted- _ J.1-__M _ ta s iei 1 TSTF-110 Incorporated 3.1-10 Rev. 2. URC afPro M 3.1-oo4 TSTF-136 Incorporated 3.1-9,3.1-15 (yaRcQ M tvm.11-cor. E TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoff l date TSTF-142 N:: [yJ. k2 -NA-- T = c;;;r;.. ;d ;;;r # ** ^ IncorperareA _ 3. t - 22. _ m_ WPt agre_ - b d -}3 1 rsrt /nc / 3Jc772 [G 3. t - 89 l (Yv$_ $ / JnicWpou4ntf / f X1-J6/ /.' K-LCL&lR2 I-Wl S/15197

CHANGE 1 NUMER JUSTIFICATION 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 prcvide adequate protection of the public health and safety. Therefore, moving of this detail is acceptable and is consistent with traveler T5TF 110.@ rAU-coM 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 l ascertained by this method. 3.1 13 This change adds an LCO 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 LCO requirement and SR were added i 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 LCO or surveillance requirement. The Surveillance Frequency of I hour is retained _from the current TS. This change is based on traveler TSTF 14ptfylgog9 re as-oos l l 3.1 14 Not used. 3.1-15 Consistent with TSTF-12. Revision 1. LCOs 3.1.9 and 3.1.11 are deleted. The physics tests contained in LCO 3.1.9 were only F contained in some plant initial plant startup testing programs. The physics test can be deleted since these physics tests are I 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 L verification. Since the N 1 measurement technique is no longer i used, the SDM test exception can be deleted. This change and l traveler TSTF 136 renumbers ITS 3.1.10 to 3.1.8. 3.1-16 This change adds the requirement to perform SR 3.2.1.2 in i 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 the F (Z) is within its limit. F (Z) is approximated by F/(Z) a a (which is obtained via SR 3.2.1.1) and F/(Z) (which is obtained j via SR 3.2.1.2). Thus both F/(Z) and F/(Z) must be established to v_erify F (Z). This change is consistent wit @pferefA10E19 a [Q 3.1 1.$ y WCGS-Differencesfrom NUREG-1431-ITS 3.1 3 S/158 7

~ CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 rese3 r3 SECTION 3.1 DIFFERENCE FRON NLREG-1431 APPLICABILITY It2EER DESCRIPTION DIAELO CANYON CONANCE PEAK' 50LF CREEK CALLAldA.Y , -i n s.s-oosj 3.1-13 In accordance with traveler TSTF-14, , the LCO and Yes Yes Yes-Yes Surveillance Kequirements are modifi o verify that thermal power 5 5% RTP. This provides an LCO requirement to correspond to CONDITION B which requires INERMAL POWER to be within limit. l 3.1 14 Not used. r R 1.I e l N/A N/A N/A N/A I I 3.1-15 In accordance with traveler TSTF-12. this change Yes Yes Yes Yes would delete ISTS LCOs 3.1.9 and 3.1.11. is change and TSTF 136 renumbers ISTS 3.1.10 to ISTS 3.1.8. 3.1 16 This change adds the requirement to perform SR 3.2.1.2 in Yes Yes Yes Yes addition to SR 3.2.1.1 during performance _of ITS 3.1.4 Required Action B.2.4.(ge@tgepVwitbtfawtGef 6 --f938-NI ) 3.1-17 Consistent with current TS LCO 3.1.3.2 and the wording of Yes Yes Yes Yes ITS 3.1.7 Conditions A and B. ITS 3.1.7 Condition C is clairified to state that the inoperable position indicators are incperable ORPI's. 3.1-18 A MODE change restriction has been adM to ITS 3.1'.1. in Yes Yes Yes Yes i the LCO Applicability, per the matrix discussed in CN 1 i LS-1 of the 3.0 package. l 3.1-19 Not used. N/A N/A N/A - N/A 1 3.1-20 Consistent with current TS 3/4.10.3. " Physics Tests." ITS Yes Yes Yes Yes I LCD 3.1.8 and its Condition C and SR 3.1.8.2 are modified } .to refer to " operating" RCS loops. i L Q21 "TR 3.1 -001 ] { lH55RTG6 h-22_- =.- INSE8tT 6 rn.3.1 - 003 ' WCGS-Conversion Conparison Table-ITS3.1 5/15M7 1

i Industry Travelers Applicable to Section 3.2 i TRAVELER # STATUS DIFFERENCE # COMMENTS l TSTF-24 Not Incorporated NA Not NRC approved as of traveler cutoff date. l TSTF-95 Incorporated 3.2-06 Approved by NRC. TSTF-97 Incorporated 3.2-07 Approved by NRC. l TSTF-98, Rev.1 Incorporated 3.2-03 1 TSTF-99 Incorporated 3.2-08 Approved by NRC. I TSTF-109 Incorporated 3.2-15 Approved by NRC. TSTF-110, Rev.(h Incorporated W 3t 24) 3.2-10 l TSTF-112, Rev.1 Not Incorporated NA Not NRC approved as of j traveler cutoff date. (A r'*E P 7RSt00+ l TSTF-136 Incorporated NA TSTF-164 Incorporated 3.2-11 Applicable to CAOC only (CPSES). ~, e ~= [43.t-t.\\ Tsvg,-3, ^', Incorporated 3.2-05,-3:iM6-Willincorporate portions 1 l gr-9y s. 2.- o9 of TSTF-25. &QV/.At(confoW/,.a f M -IQ 3 t - 2 5 \\ I I

"The. nests, wA. Frqu.sae3 -fe,esst 3.2.4.2. cara cvit ed ' conucted w% +ygical pres,entahdn &rwhihd l .provick Er a pmod of tide. af G.a estata.isN m a i h e n s. ~- i CHANGE Nl#EER - ' JUSTIFICATION ~ 3.2 12 Not a e e WCfg Seetonver.afon Cau$arMonlable) l (, os iwsr.Rf g_3a j aa,.s r i 3.2 13 inis change retains the CTS for the performance or 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 l surveillance be completed. {sent c,A.:sbRQ;5.2.-4 l 3.2 14 This change retains the Wolf Creek CTS for the completion time { for_ Required Action 3.2.2 A.2. This Completion Time was approved in License Amendment 61. This change is based on.the time 1 required to reduce power, establish equilibrita conditions, and i obtain a flux map. J ] 3.2 15 This change incorporates industry traveler TSTF-109. Action A.2 { would require the QPTR be determined rather than performing a I specific surveillance because more than one surveillance can be i used to detemine 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.) ...; 7./.: fr ". 3.2.'.? E &O te ;g re _;;erf.. .a if v..D "e z :" ^'"" '--i : ; i..7,;,;rd:J These changes are 44 3.z-10 l acceptable because they clarify the ISTS regarding frequency and 4 i use of incore flux monitoring for QPTR measurement. The changes i reflect that incore detectors provide an acceptable QPTR determination during all plant conditions. i ] 3.2 16 This change would require that both transient and static Fa measurements be detennined when performed for Required Actions c 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that is within its limit. Fo(Z) is approximated by Fa(Z) (which is obtained via SR 3.2.1.1) and F"a(Z) (which is iI obtained via SR 3.2.1.2). Thus, both P (Z) and F"a(Z) must be l established _to verify F (Z). This change is consistent with a e NMM b'f 3.2 17 br yr remen or perfo ng F,me'asur ts b a revi to orm to which not specify a omplet T C ent pr tice i o perfo measurement as s as racti . The S SR Comp ion Times are sed o what sa no ly rea ble Coup ion Time for formi a fl map:

however, problems cur, the plant be f ed t reduc i

power o shutdown. iswould'subj the nt to tran ent cond on witho sufficignf safe basis There re, ma' taining current TS requ ement s accep ble ause i M M. f k ee s.2 -C WCGS-Differencesfrom NUREG-1431 - ITS 3.2 3 5/1S/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-6 APPLICABILITY: DC, CP, WC, CA 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 QPTR," 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 at the June 1998 WOG MERITS Mini-Group meeting. These revisions corrected errors made during the development of TSTF-241. 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.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. FLOG RESPONSE: (supplement) The NRC and the Technical Specification Task Force have agreed upon the acceptable changes to be included in this traveler. Most of these changes were incorporated into the FLOG submittals in the initial RAI response as noted above. The final set of needed modifications have been incorporated as noted in the attached pages. ATTACHED PAGES: Attachment No. 8, CTS 3/4 2 - ITS 3.2 Encl. 5A Traveler Status page,3.2-1 Encl. 5B B 3.2-5 Encl. 6A 5 Encl. 6B 3 is

-. = Industry Travelers Applicable to Section 3.2 l l TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-24 Not Incorporated NA Not NRC approved as of traveler cutoff date. TSTF-95 Incorporated 3.2-06 Approved by NRC. l l TSTF-97 Incorporated 3.2-07 Approved by NRC. l 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. l TSTF-110, Rev.h Incorporated h b y M A Q n s.2 -co+] 3.2-10 i TSTF-112, Rev.1 Not Incorporated NA Not NRC approved as of traveler cutoff date. TSTF-136 Incorporated NA @ c m a 6'y M E } l m s.mj TSTF-164 Incorporated 3.2-11 Applicable to CAOC only (CPSES). TSTM 4*L. Rev. 3 ~ rppaft p 1431-M F45/ _, pious W O R-95, Incorporated 3.2-05,.1:24fr W' in l g od'Re9<2 s.2-os,3.2 2L _ T_ k ()V6%pf Kcpr6dstetP 34G) [q16-2s I \\ Madiftsd 23 greaL upN dMj IJ-/t"7l$8 mes@ be.tuxen TSTF - ApR4.. m

Fo(Z) (Fa Methodology) 3.2.1B

3.2 POWER DISTRIB'JTION LIMITS 3.2.18 Heat Flux Hot Channel Factor (Fa(Z)) (Fo Methodology) i l

L -LCO 3.2.18. Fa(Z). as approximated by F8(Z) and FRZ). shall be within the . limits specified in the COLR. APPLICABILITY: MODE 1. r i l ACTIONS COM)ITION REQUIRED ACTION COMPLETION TIME 1 i A. F8(Z) not within limit. A.1 Reduce THERMAL POWER 15 minutes = 11 RTP for each Fafter'exk. Fj(El M 11 F8(Z) exceeds limit. ftamirwh'en p 3,t.g,1 ale A.2 Reduce Power Range B_72 hours CE^j Neutron Flux-High trip h a pg(a) setpoints a 11 for each h,nt naiu 3 2*M lt F8(Z) exceeds limit. l Q 3.2-Q 8IE s A.3 Reduce Overpower AT 72 hours trip setpoints 211 for faf%.each.3:$($ -26 each it F{(Z) exceeds L=4 e: tee m in~+,g, limit. (g3,ggj 8lE A.4 Perform SR 3.2.1.1. Prior to increasing THERMAL POWER above the limit of Required A;; ions A.1. (continued) l WCGS-Mark-up ofNUREG-1431 -ITS 3.2 3.2-1 5/15/97 i

~ Fy(Z) (Fo Methodology) B 3.2.1 1 BASES ACTIONS u (continued) @senr~ B 3.'l-5gh.3.2-c. ) plant in an unacceptable condition for an extended period of (W.4RT53.2-F) C A 3. 7. - o o 2. } u A reduction of the Power Range Neutron Flux-High trip setpoints by a 1% for each it by which Fl(Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of-8 E hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. (NA 532-5 ) o.3.1-s-l u Reduction in the Overpower AT trip setpoints by a 1% for each 1% by j which Fl(Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in I accordance with Required Action A.I. quseer as.2-G]los.2-s] u Verification that Fj(Z) has been restored to within its limit, by performing SR 3.2.1.1 prior to increasing THERMAL POWER above the j. limit imposed by Required Action A.1, ensures that core conditions i during operation at higher power levels are consistent with safety analyses asstaptions. Iifiirtht~1nEnts,accanaEtaeritTWciMFdf tN32Gilie_WcR EWt'~56Hi!TtreinMM1iDPHimt[n_M c$lilieTt5ftlifeles@[nt.r]ecesserft6,2a11cCsyyMg@Ea@ Mr pioWut] wee (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.2 B 3.2 5 S/15M7

INSERT B 3.2-Sa 0 3.2-6 The maximum allewable power level initially determined by Required Action A.1 may be affected by subsequent determinations of F/(Z) and would require power reductions within-15 minutes of the F/(Z) determination, if necessary to comply with the decreased maximum allowable power level. Decreases in F/(Z) would allow increasing the maximum allowable power level and increasing power up to this revised limit. INSERT B 3.2-5b Q 3.2 6 The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of F/(Z) and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours of the F/(Z) determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. INSERT B 3.2-5c 0 3.2-6 The maximum allowable Overpower AT trip setpoints initially determined by Required Action A.4 may be affected by subsequent determinations of F/(Z) and would require Overpower AT trip setpoint reductions within 72 hours of the F/(Z) determination, if necessary to comply with the decreased maximum allowable Overpower AT trip setpoints. Decreases in F/(Z) would allow increasing the i maximum Overpower AT trip setpoints.

4 CHANGE NUMBER JUSTIFICATION {Q 3,2 -(, 95 o rifi the R ir ction y ma a. 1 1. ear at a r ucti is r 1 red thin hours L ach TR da in on, addit n, t cha exte s the i l IdO(95 ept achiev g equ bri plan ndi ~ ns to pea fact measur.nts r uir to rfo dur any power i ases fp.) owing rr ucti s requ red by QPTRJ actions. e ~ iiiksWoIffEeks ific mod ies SR 3 2.3.1 t resolve) [ liter 3.2 19 comp 1 ce c rn regar ng the of t C0 Note o yt Note st s that or mor xcore c nels must icat out of mits be e the L is not ver. 3.0.1 st s that f ure to t a SR s 1 be ) fail to meet LCO. F ' ure to SR 3.2.3 in IST d d occur i nly one core cha were ino rable. nce T the poten - confli between t te and SR. s cha is ac table se it maint s the in nt of the c ant J TS the I for AFD whi elir.inati g a lit al compl nce c"d*- q+ uw, p,. e.9 1 Not a ff C.3A to WU.fS. Sec. Con wsim Comy m M 5 3,y,.2o Tabh. ( Enclost,ue GM. heed 6A- -j Q 3. 2 r. j

3. 2 - 2.)

1 i t WCGS-Differencesfrom NUREG-1431 - ITS 3.2 5 W1587

s: 1 ' INSERT 6A-5a 'O 3.2-6 LCO 3.2.1'is revised to clarify the Completion Times consistent with TSTF-241. L The proposed change will require Actions A.1. A.2 and A.3 to be repeated after. C j. each-. subsequent Fa(Z) determination if Fa((Z) is not.within limit. This will j ensure that Actions are not. continued until the parameter is within its limit. f t: 1 i: F J I \\ l l-l [ 0 l, 1 l

4 -1 CONVERSION COMPARISON TABLE - NUREG-1431 IMPROVED TS SECTION 3.2 Page 3 of 8 t DIFFERENCE FRON NUREG-1431. REV. 1 APPLICABILITY NUPBER DESCRIPTION. DIABLO CANYON. CONANCHE PEAK WOLF CREEK CALUNAY l I 3.2-17 Fi[ [requ ~ 4ee W [ d"dttlY_^ - b % 0 0.2-12 F ^ 5 % 0 0.2-12 m Wee-.J e has revi o t ify a let Ti. & usaA. .n_N k w L 3.2-18 %1s change modifies the QP1R requirements in IMIEG-1431.3 Ilo leo Yes No Rev.1. for Wolf Creelt, to retain eemeTUrrent TS$ I requirements M '- _ m.i.. ;.;... O ^ gd 4,3,g,g g y ^:^i. ; r. m.,.i l,i ^.. le C %. appre M % [ Ame nd med M e' li fe. [ s 3.2-19 rTnty.AftIlf C spec c fi 3. .1 40e- >JA er-eJ A Vet M A -4h> w A resolve iter 31 1 t khe floteAL modi M Lawl. ps.g.q } [ t 3.2 2o -( o 3.2.-1] G,sa:ca Q cm.2-u\\ l

s. w l

1 i l r E WCGS-Conversion Cor^ _ * :: TnNe-ITS3.2 5/UM7 l t

INSERT 6B-3a 0 3.2-1 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.2-20 This CPSES specific change (CAOC plant No Yes No No only) incorporates TSTF-112. (i.e.. deletes the Note in Condition D which required that required action D.1 be completed whenever Condition D is entered). INSERT 6B-3b 0 3.2-6 i TECH-SPEC CHANGE APPLICABILITY _hUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.2-21 LCO 3.2-1 is revised to clarify the Yes Yes Yes Yes Completion Times consistent with TSTF-241. I

ADDITIONAL INFORMATION COVER SHEET-AD3lTIONAL INFORMATION NO: TR-3.3-007 APPLICABILITY: CA, CP, DC, WC REQUEST: Incorporate traveler TSTF-311 as indicated on the attached pages. ATTACHED PAGES: Attachment No. 9, CTS 3/4.3 - ITS 3.3 Encl. 2 3-10 Encl.3A 6,14 Encl. 38 6,13 - Encl. 5A Traveler Status page, 3.3-15 Encl. 5B B 3.3-64 Encl.6A 16 Encl. 6B 23 I 4 I i l.

TABLE 4.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION SURS/EILLANCE REQUIREMENTS TRIP ANALOG ACTUATING N EOR CHANNEL DEVICE WHIGH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVE4LLANCE FUNCTIONAL UNIT CHECK Cal mRATION TEST TEST LOGIC TEST SS-REQUIRED l-44-A

13. Steam Generator Water t.evel-S R21)\\

Q(15) NA NA 4r3 i - _ I~23~^ Lw-La l 1-2 -A :

14. Undervoltage - Reactor N.A.

R 21) NA O (10) N.A. 4 Coolant Pumps I-lGLS40S I-23'A

15. Underfrequency-Reactor N.A.

R 21) NA (10) NA 4 bi Coolant Pumps lil6-LS-40j

16. Turbine Trip

a. Low Fluid Oil Pressure N.A.

Ri 21) NA 10) N.A. 4 1-23-A.j

b. Turbine Stop Valve NA Ri21)

NA 10) NA 4 l-23-A M Closure

17. Safety injection input from N.A.

N.A. NA R NA 4r3 ESF OM QQ

18. Reactor Trip System Interlocks l 9 I-U l

a. Intermediate Range Neutron flux, P-6 NA R(4)

R N.A. NA 3##

b. Power Range Neutron Flux, P-8 N.A.

R(4) R N.A. NA 4

c. Power Range Neutron Flux, P-9 N.A.

R(4) R NA NA 4 l lo exceedm3 he. P-s efectock haver t.7g 3,g g 1 -me, unt4-h>s been in @E S H OLF CREEK-UNITI 3M 3-10 Amendment No.43 43 7 Mark-up ofCTS3M. 3 5/15M

CHANGE NUPBER NSHC DESCRIPTION i an additional hour is provided for the transition to H00E 3. 1 14 A In the ISTS. Table 3.3.11 Function 20, the Reactor Trip Breaker (RTB) Undervoltage and Shunt Trip Hechanisms are separate from the RTB Functional Unit. The current TS have been revised to reflect these requirements. new [ footnote (b)] has been added to the RTB Functional Unit to rote that the same OPERABILITY requirements and ACTIONS tpply to a bypass breaker if it is racked in and closed for bypassing an RTB. The bypass breakers j were already handles in this fashion. Action Statement [12] in current T le 3.3 1 has been revised accordingly, usrr.cr.sA -r,4 l4i4l __ _ _ 1 15 A The Applicability for the Reactor Trip on Turbine Trip h h A h c.a M t(c. Q o M function is modified by a new footnote [(a)] such that reb.t td A bs. n.vM TAM this function is only required to be OPERABLE above the p u.aw3m'I A 4M-P 9 interlock setpoint (50% RTPL This h acceptable since the trio functio ocked below P 9.'[New] ,43.3421 g, J ACTIONStatement[1$ ied to the Low Fluid Oil Pressure and Turbine Stop Valve Closure trip functions. 1 These are consistent with NUREG 1431. Rev.1. 1 16 LS 40 The requirement to verify the setpoint during the quarterly TADOT for RCP Underfrequency [and RCP Undervoltage] is deleted, consistent with NUREG 1431. Rev. 1. 1 17 A Consistent with NUREG 14?1 Rev. 1, 3.3.1 Requi Action D and E Note, the current TS Table 3.3-L ACTION Statement 2 goe hey m,pCRJtitsmFnt.2.'7ASv$been l 4 3 3-40 modified by a NOTE that allows the bypass to be used" for surveillance testing or setpoint adjustment. Setpoint adjustment can be performed at power and may be required by other Technical Specifications. The { reason for placing the channel in bypass does not effect the impact of having the channel in bypass. 1 18 LS 7 The current TS requirement to reduce the Power Range l Neutron Flux Trip setpoint in the event a power range l-flux channel is inoperable, is deleted. This deletion L is consistent with NUREG 1431. Rev.1 and is justified by: 1) The loss of one channel does not impact the reliability of the Reactor Trip System because l the affected channel is placed in trip. It may, WCGS-Description ofchanges to CTS 3M.3 6 S/15/97

CHANGE NUMBER NSBC DESCR1PTION 1 54 LS 37 Not applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 3B). 1 55 LS 39 Applicability Note [*] and ACTION Statements [5.a and 10] for Functional Units [1, 6.b. 19. and 20] of cuarent TS Table 3.31 are modified to provide an alternative to opening the reactor trip breakers (RTBs) while still assuring that the function and intent of opening the RTBs is met. As currently worded, these ACTION Statements result in a feedwater isolation signal (FWIS) when in H0DE 3 with a T,, less than [564'F. USAR Table 7.3 14 and USAR Figure 7.2 1 g (sht.13) detail the FWIS generation on the coincidence 1 / reqdN M of P 4 and low T.,.] A more generic action, which ctctik be sM assurehhe rods are fully inserted and cannot be withdrawn, replaces the specific method of precluding l pa.ts-rsed t rod withdrawal. The revised Applicability and ACTION }TR 3.3 M I Statements still assure rod withdrawal is precluded. This change does not involve any safety impact and is consistent with traveler. 1 56 Not applicable to Wolf Creek. See Conversion W 1-s4 Comparison Table (Enclosure 3B). 1 57 LG Current TS Table 3.31 Functional Units [12.a and 12.b] are combined per traveler TSTF 169. The Required Channels, ACTION Statement, and Surveillance Requirements are the same for both Functional Units. The only difference between the two is the Applicability which could lead to entry into ACTION Statement 6 for Functional Unit [12.a]. followed by a power reduction below P 8 exiting the Applicability and required actions for that Functional Unit and l subsequent re entry in ACTION Statement 6 for Functional Unit [12.b]. This would involve an improper etnulative A0T of 12 hours before tripping an I inoperable channel, beyond that evlauated in WCAP 10271 and its Supplements. The relationships between these Functional Units and permissives P 7 and P-8 are moved to the ITS 3.3.1 Bases. 1 58 A Not applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 3B). @ s,.t ppu cMa. b Wif. Creek.. S= G-niA, 1 59 Ne

c. _,

.m%u._ sy ..e o 1 60 Not.usesk appii cabia +u QM Cre ek.. see Gov ** Cbmpm A wT.Ma. C Ehc.L>mni '3 G 7 [NW k 1-61 M Not applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 3B). =_ I hs&RT 3A - M)- WCGS-Description ofchanges to CTS 3N.3 14 S/1S/97

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.3 P.ne 6 or3e TECil SPEC CHANGE AFFLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALIAWAY 1-15 The Applicability for Reactor Trip on Turbine Trip function No - see No - already YM Yes A is modified by new footnote [(a)] such that this function CN 1-48 LS in current M [11e um ( is only required to be OPERABLE above the P-9 interlock Lalb u t-6_- 9 o jo s.3-o2d setpoint (50% RTP). This is acceptable since the trip (g,3,g g3,gj n],_ - - - gm g -g $ applied to the Low Fluid 01([New] ACTION Statements function is blocked below P 9. T & d E8 \\

513 Applical4IQ ab Pressure and Turbine Stop Valve Closure trip functions.

,,.,h. m,4eA. rm ocrr r y % MT4els 4.51.) 1-16 The requirement to verify the setpoint during the quarterly Yes Yes Yes Yes LS-40 TADOT for RCP Underfrequency [and RCP Undervoltage] is deleted. 1 17 The bypass allowance can be used for surveillance testing Yes Yes Yes Yes A or setpoint adjustment. Setpoint adjustment can be performed at power and may be required by other Technical Specifications. 1-18 The current TS requirement to reduce the Power Range Yes Yes Yes Yes LS 7 Neutron Flux Trip setpoints in the event a power range flux channel is inoperable, is deleted. The time to reduce power below 75% RTP is increased from 4 hours to 12 hours and, if actions are not completed as_ required, t_he unit must be in MODE 3 within 12 hours. { y g pp 4 3.t120 } 1-19 If the requirements of current ACTION Statement [6] are not Yes No - see Yes Yes LS-8 met. LCO 3.0.3 would be entered. In accordance with the CN 1-61-H ISTS. this ACTION Statement is revised to state that, if the ACTION requirements are not met, the plant must be taken below the P-7 interlock setpoint within the next 6 hours. [The Applicability for Functional Units 9,11.12,

14. and 15 in current TS Table 3.3-1 is also revised to add new footnote (c)].

i S/1587 WCGS-CONVERSION TABLE-CTS 3/4.3

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.3 rage is orso TECII SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 1-55 Applicability Note [*] and ACTION Statements [5.a and 10] Yes Yes Yes Yes LS-39 for Functional Units [1. 6.b. 19 and 20] of current TS Table 3.3-1 are modified to provide an alternative to opening the reactor trip breakers (RTBs) while still assuring that the function and intent of opening the RTBs is met. 1 56 The DCPP CTS 3.3.1 Action 2.c requires that power be Yes No No No reduced to then 75% or that SR 4.2.4.2 be performed py whenever power is t 50%. This power level requirement g, should be 2 75% since if power is decreased below 75% per the first part of Action 2.c. the required Action is complete and in addition SR 4.2.4.2 is only required for power levels t 75% with one power range detector inoperable. 1 57 Current TS Table 3.3-1 Functional Units [12.a and 12.b] are Yes Yes Yes Yes LG combined per TSTF - 169. The relationship between the Functional Units is moved to the Bases. 7 1-58 The proposed change would allow Reactor Trip System and No - see Yes No - see No - see i A ESFAS sensor response time testing to performed per WCAP-CN 1-03-LS-1. CN 1-03 LS-1. CN 1-03 LS 1. 13632 P-A Revision 2. " Elimination of Pressure Sensor Response Time Testing Requirements " or other similar methodologies. This change is consistent with traveler TSTF-111, which revises the Bases for ITS SR 3.3.1.16 and SR 3.3.2.10 to allow the elimination of pressure sensor response time testing. hv4 3 h _lOl-Stj 1-59 (used.) hsev+ 38-133 h h h @3 3 @~I 1-60 WCGS-CONVERSION TABLE-CTS 3M.3 5/1587 4

INSERT 3B-13a TECH SPEC CHANGE APPLICABILITY-l NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 1-59 The CTS for DCPP currently requires that Yes No No No l-LS-46 interlocks 'P-7, P-10, and P-13 be surveillance tested via a CHANNEL CALIBRATION and a COT each refueling Q1-51} I outage. This change deletes that requirement for P-7 and substitutes an ACTUATION LOGIC TEST 1-60 The TAD 0T Frequency for Functional Units No - see CN 1-Yes No - see CN 1-No - see CN 1- + A 16.a and 16.b in CPSES CTS Table 4.3-1 is 48-LS-4 15-A 15-A revised to be consistent ~with the " MODES .TR. S.3-007 l For Which Surveillance is Required" column. i I b + wm

_ _ _.. _ _ ~. _ _ _ _. _... _. l I INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.3 E i TRAVEIIR # STATUS DIFFERENCE # COMMENTS (weese.hs p m h TSTF-19, Rev.1 @ncorporated N/A $W W A 3 MD VQ S.3-34 \\ TSTF-37, Rev.1 Not Incorporated NA ITS 5.6.8 still addresses PAM reports. Sections afterITS 5.6.7 were not renumbered. TSTF-51 Not Incorporated NA Requires plant-speciSc reanalysis to establish decay time dependence for fuel handling accident. [T8TFA Dieflee6tpotjned!-- [ a } Nk b fo M s.3-o*SI J,1 yo ,e in - e TSTF-111 RevhIncorporated NA W t-osJ TSTF-13 hl 33-41,h /frav ~ r i7s h Incorporated 3 3-93 iin stur ho d h. e bee 3 3-95 eve sep te vele. l TR 3.1-oot. 33- ,Po ens f the ve that! S L4't [ gn' ant al as.so o bili requ e ij.iM e b n inc o ed. TSTF-161[lw] Incorporated 3 3-79 Q@ TSTF-168 Incorporated 3 3.43 [ a b [ A ]r le85-'81 TSTF-169 prporated 33-42 [perekbj afE] fra u-omj l (W6Er-J8[6 Ncorporated 3 3-49 Ie 1s-W ~ S-M ro Incorporated 33-107 MAetb.atetim3) [ cps.3-107] e STF - 31 % lncorpratecL, 33 (TR,5.3 -cD 1 \\ 5/15/97

RTS Instrunentation 3.3.1 l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.12 NotTUsJd! ) ~ $$1SSR{ n,e .~,, Thi; Suruilla;; sh:11 irclud; ;;rific; tier. of i ". c;ter 0;;l at Sy;t s r;;i;t m a tu.g r;tur; I d; tater byp;;; 13;p ;w retc. "erfers. 0".".5L CALI","AT!". le-menthe gggggg 1 4 4 SR 3.3.1.13 Perform COT. 18 months SR 3.3.1.14 - ------- - - - NOTE Verification of setpoint is not required. Perform TADOT. 1_8 months teat.8%pt's pwm 3-SR 3.3.1.15 NOTE - - 3,sys Verification of setpoint is not required. O r .1 re w t 4 i ) rf w nl vio d s I .....q Perform TADOT. P or to ctor i \\ start N s (continued) Prso,r to exceedq he-P.S inte. clock, wheneveribe. [ wet ha bee.n m t10DE S, 'IS33* col ( if nq per{wmed. G h (per.viou.s a t c49_. - WCGS-Mark-up ofh71 REG-1431 - ITS 3.3 3.3-15 S/15R7

7 ~ __ _ _ _ _ __ _ _ _ __. _ _ _ _ _._. i RTS Instrtmentation B 3.3.1 BASES ~~ = 'eaed +he. P-f shteriock whenevw SURVEILLANCE SR 3.3.1.15 he wN hu bun W M00E 3. "Th R 3.Le07 REQUIREMENTS 4 t (continued) SR 3.3.1.15 is the performance a TADOT of Turbine Trip Functions. This TAD 0T is as cribed in SR 3.3.1.4 except that I this test is performed prior to h Anosrstates (fiiFS@ Surveillance is not required if it has been performed within the previous 31 days. Verification of the Trip Setpoint does not have to be performed for this Surveillance. Performance of this test will ensure that the turbine trip Function is ) OPERABLE prior td ng r or r1 al Thi es .a t be rf i act at r nd st t ref e rf riy to act st up. @ceedh the P-9 sh+< dock.) \\ " * ' SR 3.3.1.16 {Q"a3-56 \\ SR 3.3.1.16 verifies that theMindividual channel Are4e actuation response times are less than or equal to the maximum 3_3 g~g* values assumed in the accident analysis. Response time testing g3,3 55 j accepta iteria are included in Ts.'...ini.";girc. c.t; Manuel-Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parrmeter exceeds the trip. setpoint value at the sensor t; tra p.nt ;t Jhich tt;

wi-- 7.t racta; tra r; wired fur.;ti;ral ;t;t; (i.e., estr;l sd ;hatta r;t fully iracrt;d in th; renter wre)Lintf17eiss oEdstionaryatippercoi]M1tage.

For channels that include dynamic transfer Functions (e.g., lag, lead / lag, rate / lag, etc.), the response time veriffcatiortTstest may-be performed with the tra;f;r rurati;n ;1Rm ;;.;i at te era C., aith th; rsultin; =nur;d rap.x tim c pred t; tte ;pprepri;t; I';?", r;:-;-:-7.x tin. Alter ately the ra p ax tin tat Gn be perfered Jith the tia G.at;at; ;;t t; th;ir . W ral v;1 x;. previ id the r; @ ired rc pr.x ti n i;

rel3tially Gl;uleted n;;ing th; time constants are set at their nominal values. The response time may be measured by a ries of overlapping testsCoriotherjerffilcatioif{eirg.T.~ rec 7

such that the entire response time is measured. lo s.s.ss[ RespenseitimOnay;[belverifiedf byfactuallresponsei. tigtestslin arjEeriesJf.fsequentia]Mwer, lapping;oritotaEctinnel measurements'.SorJby the' summation of allocated response times-4 1 \\ (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.3 B 3.3 64 5/15197

CHANGE NlhEER JUSTIFICATION 3.3 125 ITS SR 3.3.1.11 is modified by a Note that requires verification that the time constants are adjusted to the prescribed values. The addition of this Note is consistent with SR 3.3.1.10 and is required because SR 3.3.1.11 is used for the Power Range Neutron Flux - High Po.sitive Rate [] trip function whi:h has a time constant associated with its calibration. 3.3 126 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). 3.3 127 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). I 3.3-12B Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). 3.3 129 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). 3.3 130 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). 3.3 131 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). 3.3 132 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). 3.3 l'43 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). IWC 3.LoQ 3.3 134 The in)cfndit K o dTS 3 .2 ha en is onsi ent3 w curr it TS able .3 ion [ for ncti 7b n t 43 pro 'de 4 urs f the add ona han to be aced bypesfIfor rveil nce t ting oth ch4 h e NseRT (4 A-wo ~ 3.3 135 A MODE change restriction has been added to ITS 3.3.1 Condition C l per the matrix discussed in CN 102 LS-1 of the 3.0 package. (See the LS-1 NSHC in the CTS Section 3/4.0, ITS Section 3.0 package). 3.3 136 The TAD 0T performed under ITS SR 3.3.2.7 includs verification of relay setpoints since the trip actuating devices being tested are the same circuits tested under ITS SR 3.3.5.2. [th!S&AT GA-ILl WCGS-Dfferencesfrom NUREG-1431-ITS 3.3 16 5/1S/97 l AJ sGET 6 A - IGk "c 3 3 -# 9 j 3 -14 5 i

l I INSERT 6A-16 CHANGE HMMBER JUSTIFICATION j \\c41s.ayll 3.3-147 Not Applicable to Wolf Creek. See Conversion-Comparison Table (Enclosure 68). ICP13.aNE l 3.3-138 Not Applicable to Wolf Creek. See Conversion Comparison Table ~ (Enclosure 6B). IDc-Au..co2d 3.3-139 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). 1 lcA 3.Sco1 1 1 3.3-140 Not Applicable to Wolf Creek. See Conversion Comparison ~ Table (Enclosure 68). lcA3.3 col \\ 3.3-141 Not Applicable to Wolf Creek. See Conversion Comparison laDie (Enclosure 68). 43.ts.eru ' l T E 3. 3-406e_ 3.3-142 Adds "or trains" to ITS Condition A. consistent with ITS 3.3.2 and traveler TSTF-135. l rR11-col.l 3.3-143 The Frequency of ITS SR 3.3.1.15 is revised to reflect traveler TSTF-311. The Applicability of these Turbine Trip Functions 1 begins in MODE 1 at the P-9 In+.erlock (50% RTP). The STS 1 Frequency. " Prior to reactor startup." is inconsistent with the Applicability of these' Functions which can be tested at power with minimal perturbations to plant systems. These TADOTs are typically performed prior to loading the main turbine, if not performed within within the previous 31 days. Both Callaway and Wolf Creek have filed LERs [ Wolf Creek LER 97-022 00. Letter WO 97-0139 dated December 22, 199?f regarding this issue. This change was previously approved for Vogcle. I Q3.3-144T 3.3-144 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B), i

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 23 of 23 SECTION 3.3 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.3-134 This change is Wolf Creek specific to revise the h0TE in No No Yes No - see CN 3.3-18. Condition K of ITS 3.3.2 consistent with current TS Table + -a { gggs 1 1 tion ~for tion 7 and f --. 4J 3 3 -3. A'it'" 30 f*' I"*di J l f p.3.3 N 'l ide rs an o . pl n / gypa or s eilla_ esti of o r_ch is 3.3-135 A MODE change restriction has been added to ITS 3.3.1 Yes Yes Yes Yes Condition C per the matrix discussed in CN 1-02 LS-1 of

the 3.0 package.

3.3-136 The TADOT performed under ITS SR 3.3.2.7 includes No - adopted ISTS No - adopted ISTS Yes Yes verification of relay setpoints since the trip actuating format. format. devices being tested are the same circuits tested under ITS SR 3.3.5.2. tusant G B-23 L 3--145 14s.r2.T G13 -23 e hc_3.vot9 \\ i i WCGS-Conversion Comparison Table-ITS3.3 S/2S/97

INSERT 6B-23 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.3-The Condition for Function 4.c is changed Yes No - Insert 1 No - Insert 1 No - Insert 137 from Condition D to E consistent with DCPP !O^ M# I CTS 3.3-This CPSES-specific change revises Table No Yes No No 138 3.3.1-1. Note 1 to reflect new Overtemperature N-16 parameters approve for Unit 2 in Amendment 55/41, and submitted (cP3.3-oo3l for Unit 1 in LAR 97-003. 3.3-This change adds new SR 3.3.2.13 which is Yes No No No 139 the performance of an 18 month TADOT. SR %mm2-) 3.3.28 is the performance of a TAD 0T every 24 months for DCPP. 3.3-Changes to RTS and ESFAS delta T Functions No No No Yes 140 are made to reflect Callaway OL Amendment gn, No. 125 dated 4/13/98. 3.3-Changes to ESFAS Feedwater Isolation No No No Yes 141 Functions are made to reflect Callaway OL %A 3.S-cc2.] Amendment No. 126 dated 4/23/98. 3.3-Adds "or trains" to ITS Condition A. Yes Yes Yes Yes o yt.s-c,s.O 142 Consistent with ITS 3.3.2 and TSTF-135. TW.1M A [ den-

A TsTs:-$_1LNet. art.st. "Rev the. SR. 3.3.1. 6 %mg W ha.

~' NAfr Yes Mr Yes Mk s 3.3-a 3.5-06T 143 3.3-SR 3.3.7.3 and 3.3.7.4 are deleted for DCPP Yes No No No 144 since there are no actuation logic or master relays associated with the CRVS pressurization system actuation via the --l Q 3.3-lW CRVS atmosphere intake radiation monitors. Insert 1: Bypass Action applies to containment pressure High-3 not High-2 (different design). t a b t. .-m

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.3-019 APPLICABILITY: WC, CP, CA, DC REQUEST: In the conversion from CTS to ITS, information contained within Action Statements that modified the initial required action was converted as a Note. Based on formatting requirements of NUMARC 93-03, " Writer's Guide for the Restructured Technical Specifications," the span of a Note is limited to the width of the text it modifies. Two case types currently exist within the ITS which require revision in order to meet both the intent of CTS requirements and formatting requirements of NUMARC 93-03. In the first case, notes span the full width of the Required Action column when they should only apply to a single Required Action. Notes that fall within this category and require reformatting to conform to both CTS and NUMARC 93-03. In the second case, notes span only a single R6 quired Act.'on when they should apply to all Required Actions for the associated Condition. Notes that fall within this category and require reformatting to conform to both CTS and NUMARC 93-

03. It was not the intent of the CTS that the clock for shutdown actions should start when a channel was bypassed for testing as allowed by these Notes.

ATTACHED PAGES: Attachment No. 9, CTS 3/4.3 - ITS 3.3 Encl. 5A 3.3-3,3.3-26,3.3-27,3.3-28,3.3-29,3.3-30 Encl. 6A 16 Encl.6B 23 l Q

RTS Instrumentation 3.3.1 i ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. One Power Range NOTE Neutron Flux-High The inoperable channel may be channel inoperable. bypassed for up to 4 hours for surveillance testing and setpoint adjustment of other channels. D.1.1 Place channel in trip. 6 hours M D.1.2 Reduce THERMAL POWER to 12 hours l s 75% RTP. i E l D.2.1 Place channel in trip. 6 hours beat "... f..s.M.......N0TE l Orb @ reouired to be gu,g performed WWMyytey' n " F N 7 J a. thq Power .gd. (q'j i Range Neutron Flux 6.4eput+-> to OPTR is inoperable.g (3.b-ME) NMX+75513P] I W J3*9 l h Perform SR 3.2.4.2. Once per 12 hours l E D.3 Be in H0DE 3. 12 hours i l (continued) 4 WCGS-Mark-up ofNUREG.H31 -ITS 3.3 3.3-3 S/15/97 i - l

l 1 ESFAS Instrumentaticn 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One trair! inoperable. h ....- NOTE -

ggggy4 F

" One train may be bypassed c.i g for up to 4 hours for b3-b D9. surveillance testing Ltx u oti) .3.a..(2-) in in o F, g - provided the other train Enter-ar@c,6 Mba s is CPERABLE. ,,x g m p M Ac.+\\m,s of ,,,,j,g y 3.33 tco 3.6 3, wtam nd Restore train to OPERABLE 6 hours 143.2 32.\\ tsolaisen Valves,,, k cet" o g l status i P W"M'#" ##\\"'* b ? isopuc. h3 ialah'M g in sinuwuM>{t'm, 3 Be in MODE 3. 12 hours Cf2 Be in MODE 5. 42 hours ) h he inoperable channel D. One channel c

NOTE inoperable, e

T Oy:gy gi bj k may be bypassed for up to 3,3,,4g 4 hours for surveillance testing of other twc.s.3-ots } channels. Place channel in trip. 6 hours E D.2.1 Be in MODE 3. 12 hours 6NQ D.2.2 Be in MODE 4. 18 hours (continued) 4 WCGe riark-up ofNUREG-1431 -ITS 3.3 3.3 26 S/15/97

. ~. - -.. ESFAS Instrumentation 3.3.2 ACTIONS (continued). CONDITION REQUIRED ACTION COMPLETION TIME h NOTE E. One Containment Pressure channel e One additional channel ggggggg Mt inoperable. "" may be bypassed for up to c 4 hours for surveillance 33-H6 .I................ \\" 33*S l Place channel in bypass. 6 hours 2 E.2.1 Be in MODE 3. 12 hours E.2.2 Be in MODE 4. 18 hours F. One channel or train F.1 Restore channel or train 48 hours inoperable. to OPERABLE status. E F.2.1 Be in MODE 3. 54 hours M F.2.2 Be in MODE 4. 60 hours (continued) ' WCGS-Mark-up ofNUREG-1431 - ITS 3.3 3.3 27 S/1S/97

i-t i ESFAS Instrumentation l. 3.3.2 1. ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME g G. One' train inoperable. h ! -- - - NOTE - gggg One train may be bypassed ~ " ' ' A__g for up to 4 hours for I3-14E l 4db surveillance testing l provided the other train I."5 M is OPERABLE. 1 l Restore train to OPERABLE 6 hours statas. E G.2.1 Be in H00E 3. 12 hours 8E } G.2.2 Be in MODE 4. 18 hours h - - train may be bypassed H. One train inoperable. NOTE---- --- - l One ggggj M for up to 4 hours for 4% surveillance testing 3 3'M provided the other train is OPERABLE. [as.S oM j L........................ i\\ Restore train to OPERABLE 6 hours

c.

status. i E H.2 Be in H0DE 3. 12 hours 1 i (continued) i l l WCGS-Mark-alp ofNUREG-1431 - 1TS 3.3 3.3 28 $/1587 l

ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME h ...- NGTE - - --- - I. One channel inoperable. [ ~The inoperable channel D' g may be bypassed for up to p.3,93} 4 hours for surveillance testing of other channels. y Place channel in trip. 6 hours t E I.2 Be in MODE 3. 12 hours J. One Main Feedwater -.g? r 3-NOTE B~ffiF~9; i Pumps trip channel Tht.1glperilitDifjannerapyJR 7 3';32569 inoperable, tantlnEd*futJD;'t0Th?sifDr. Smytfl14nctNstinj[orother chamelo ^iES i + iE:.- :3?L - -f-- J.1 RestorePlacechannelto ^^t"J."L: ;t;tus in._ trip. 1 40 hours E J.2 Be in H0DE 3. 7 54 hours (continued) IVCGS-Mark-up ofNUREG-1431 - ITS 3.3 3.3 29 S/15 8 7

ESFAS Instnmentation l 3.3.2 ACTIONS (continued) l COWITION REQUIRED ACTION COMPLETION TIME l h --- --- NOTE h L K. One channel P l inoperable. i -One additional channel E3M* I ~ = M may be for up to 4 hours surveillance g $MMIA j' Place channel in bypass. l 6 hours ggggg l K.2.1 Be in MODE 3. 12 hours M K.2.2 Be in MODE 5. 42 hours J l \\ l 4 L. One o6sibreggefred L.1 Verify interlock is in 1 hour (gg i channel [G required state for inoperable. existing unit condition. 1 08 L.2.1 Be in MODE 3. 7 hours l M l L.2.2 Be in MODE 4. 13 hours (continued) 1 e i WCGS. Mark-up ofNUREG-1431 -ITS 3.3 3.3 30 S/1587

. ~. r i l CHANGE IGBER JUSTIFICATION 3.3 125 ITS SR 3.3.1.11.is modified by a Note that requires verification that the time constants are adjusted to the prescribed values. The addition of this Note is consistent with SR 3.3.1.10 and is required because SR 3.3.1.11 is used for the Power Range Neutron Flux - High Positive Rate [] trip function which has a time l constant associated with its calibration. 3.3 126 Not Applicable to Wolf Creek. See Conversion Comparison Table -(Enclosure 68). l 3.3 127 Not Applicable to Wolf Creek. See Conversion Comparison Table p (Enclosure 68). L 3.3 128 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). 3.3 129 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 6B). l~ 3.3 130 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). 3.3 131 Not Applicable to Wolf Creek. See Conversion Comparison Table (Enclosure 68). i-i l 3.3 132 Not Applicable to Wolf Creek. See Conversion Comparison Table l. (Enclosure 6B). 3.3 133 Not Applicable to Wolf Creek. See Conversion Comparison Table L (Enclosure 68). = lWC. 3.3-ot 6 l 3.3 134 The in ndit Ko TS 3 .2 ha is onsi ent3 ~ w curr t TS able .3 ion [ for ncti 7b . u.-4 at 43 pro 'de 4 urs f the add ona han to be ced b for rveil nce t ting ot ch 1 4 e n T 6 A -it.a ~ 3.3 135 A MODE change restriction has been added to ITS 3.3.1 Condition C per the matrix discussed in CN 102 LS 1 of the 3.0 package. (See the LS 1 NSHC in the CTS Section 3/4.0. ITS Section 3.0 package). E 3.3 136 The TADOT performed under ITS SR 3.3.2.7 includes verification of relay setpoints since the trip actuating devices being tested are l the same circuits tested under ITS SR 3.3.5.2. l [!MssAT GA-IL WCGS-Differencesfrom NUREG-1431-ITS 3.3 16 S/158 7 -l45 idsEtT (4 -IGppc 3 3- '9 j

INSERT 6A-16b WC 3.3-019 '3'3-145-In converting from CTS to ITS, information contained within an Action Statement that modified the initial required action was converted to a Note. Based on formatting requirements of NUMARC 93-03, " Writer's Guide for Restructured Technical Specifications," the span of a Note is' limited to the width of the text it modifies. Two case types currently exist within the ITS which require revision in order meet both the intent of CTS and formatting requirements of NUMARC 93-03. l In the first case, notes span the width of the Required Action column, but the intent in CTS is to modify a single required action. In the second case, Notes span a single Required Action, but application of the Note within CTS, affects all Required Actions for the associated Condition. The span of affected Required Action Notes has been revised to meet both the intent of the CTS and formatting. requirements of NUMARC 93-03. l l l h l l I r l I l l 1 4

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 23 of 23 SECTION 3.3 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK OLF CREEK CALLAWAY W 3.3 134 This change is Wolf Creek specific to revise the NOTE in No No Yes No see CN 3.3-18. tion K of I 3.3.2 consistent with current TS Table * ~ Q { c1wo j N ~7-f ide ars an onal a to pl n t ypa or surfeilla_ esti of a r ch is. 3.3 135 A MODE change restriction has been added to ITS 3.3.1 Yes Yes Yes Yes Condition C per the matrix discussed in CN 1-02 LS 1 of the 3.0 package. 3.3-136 The TADOT performed under ITS SR 3.3.2.7 includes No - adopted ISTS No - adopted ISTS Yes Yes verification of relay setpoints since the trip actuating format. format. devices being tested are the same circuits tested under ITS SR 3.3.5.2. I susERT 66 ~2.~5 3 -14 5 14 se.r2T (,13 - 23 e y c_ s.s-o n ( S/15/97 WCGS-Conversion Comparison Table-ITS3.3

. ~. - _. INSERT 6B-23a WC 3.3-019 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.3-The span of affected Required Action Notes Yes Yes Yes Yes 145 has been revised to meet both the intent of the CTS and formatting requirements of NUMARC 93-03.

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.1-2 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-40 Comment: WOG-99 has not yet become a TSTF. FLOG RESPONSE: WOG-99 has been designated TSTF-282 which is currently under NRC ) review. No changes to the ITS mark-ups were made in the process of assigning this traveler a TSTF number. As explained in Enclosure 6B to 0, JFD 3.4-40 does not apply to CPSES or DCPP. Those plants are retaining their CTS, as explained under JFDs 3.4-34 and 3.4-51, respectively. Callaway and Wolf Creek continue to pursue the j changes proposed by this traveler. FLOG RESPONSE (supplement) Based on the present status of the generic change process for the STS, it appears that traveler TSTF-282 will not be approved by the NRC in time to support the initial license amendments for the FLOG plants. In order to facilitate the issuance of these initiallicense amendments, an alternate approach has been developed which relies on the CTS, plant-specific information, and/or the NUREG but does not rely upon the traveler. This alternate approach is hereby provided as an interim submittal to allow issuance of the initiallicense amendments. The changes which rely upon the traveler can be processed in subsequent license amendments following approval of the traveler by the NRC. For Callaway, CTS SR 4.2.5.2.a is retained to require RCS flow be calculated prior to exceeding 75% RTP after each refueling outage, and SR 4.2.5.4 is retained which requires RCS flow be determined by precision heat balance measurement on an 18 month frequency. SR 4.2.5.4 in not tied to the beginning of a cycle. This is acceptable because other indication of RCS flow is available (RCS flow meters) until such time that plant conditions are established which are suitable for a precision heat balance. ITS SR 3.4.1.4 and its Bases are revised (new JFD 3.4-55) to additionally require the RCS flow be calculated prior to exceeding 75% RTP after each refueling. For Wolf Creek, the Note to ITS SR 3.4.1.4 will be maintained with the 7 day interval based on CTS. CTS SR 4.2.5.1 states that the provisions of Specification 4.0.4 are not applicable to RCS flow rate. Maintaining the ITS note allows performance of the SR in MODE 1 when the plant conditirs support performance of the SR. The 7 day intervalis approp-4te based on the prerequisites and time to perform the precision heat balance. CTS 4.2.5.4 did not specify a time frame 'or completing this surveillance. Current plant practice is to perform this surveillance as soon as practical after the plant reaches 100% RTP after a refueling outage. Calibration of the instrumentation used for determining steam pressure, 1 feedwater pressure, feedwater temperature, and feedwater venturi Ap in the calorimetric calculations typically takes three shifts (approximately 36 hours). Following instrument calibrations, data collection for the precision

^ he:t b.lince requir:s approximattly two hours. Once th3 data has been collected, approximately 12 hours in necessary to complete the calculations associated with the precision heat balance. Therefore, the NUREG time interval of 24 hours for Wolf Creek is not acceptable and the 7 day interval is consistent with the intent of the CTS. i ATTACHED PAGES: !, CTS 3/4.2 - ITS 3.2 Encl. 2 - 2-15 Encl. 3A 9 Encl. 3B 7 ! 0, CTS 3/4/4 - ITS 3.4 Encl. 5A. Traveler Status page Encl.6A 7, 8 Encl. 6B 6 I 1 i l I l

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMmNG CONDITION FOR OPERATION ACTION: (Continued) d_. ^; M hrtect tM er:- d the out-of-limit >. ondition 50$0$167 4. prior to increasing THERMAL POWER above the reduced THERMAL POWER NM** limit required by ACTION 1.b and/or 3, above; subsequent POWER OPERATION may proceed provided that the indicated RCS total flow rate is demonstrated to be within the region of acceptable operation pnor to exceeding the following THERMAL POWER levels:

a. A nominal 50% of RATED THERMAL POWER,

b. A nominal 75% of RATED THERMAL POWER, and

c. Within 24 hours of attaining greater than or equal to 95%

of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.5.1 The provisions of Specification 4.0.4 are not applicasse to Specification 3.2.5.c. 4.2.5.2 Each of the parameters of Table 3.21 shall be verified to be within their limits at least once per 12 hours. DF-11. -A 4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL , - CP 3.2-00 l} CAllBRATION at least once per 18 months. W- -undstets. I" undelats.__ 4.2.5.4 The RCS total flow rate shall bobs (id;.7 '--- 'nd by pir" :^- h-g_ p.t g., t g ~.x r_:-- z- ; "at least once per 18 monthsh"M- ' fr;: a h m,g7 g j kQ3A A~L \\ ,,, g$, t

n: * ;; ';.. _..; M prrfi -- 50 ' M'-.x, S: :n^ ;m; _ :-

c t reed 8 r " - _ - - - ' '. 9 :': m ;:-- r e, ' :f:::t::; r-r- ' :f ::t:: ^ ..z-, nd ':: f ::^: nn'2d ^.P '- 'h: :': =c;.::: -f-"' rye.: f@0$4MR . 6. 5 _6...6 .. a Mu n *+M 415.5 Se ' rifrier re-'ud tha!! S ! :;::^:I'^ r%;; :M frened 2: ' ' 05-09-LG $

---- j et ':::t xx per 15 rn.

" - - - ~ h Not required to be performed until 7 days after achieving 2 95 % RTP.) g5 $ h p 3.4.1 M WOLF CREEK - UNIT 1 3/4 2-15 Amendment No. 61 Mark-up ofCTS3M.2 S/IS/97

1 CHANGE NlABER N2iG DESCRIPTION maintained within specified limits in order to ensure consistency with the assumed initial conditions of the accident analyses. The limits placed on the RCS temperature. pressure, and flow ensure that the minimum departure from l Nucleate Boiling ratio (DNBR) will be met for each of the l transients analyzed. Compliance with the above limits is verified every 12 hours. If a parameter is found to be outside the required limit. 2 hours are allowed in order to restore the parameter to within the limit. If the parameter l is not restored to compliance within the required time, the plant must be shut down. The revised completion time of 6 hours is acceptable to allow transition to the required plant conditions in an orderly manner without unnecessarily initiating any undue plant transients and on the small likelihood of a severe event occurring during the extended time period. h This M rvei anchor 05 07 as ing S fl by pr isio heat ba nce " 1ed o a tnot that rres. s to te f IT SR 3 .1 foo te r uires lat t sur ill e o with 7d of a ieving

RTP.

s mor es ict e in at i ties surv lance t i t i ng fa ycle. i s acc able ause o r i ica on RC flow s av able CS fl ters nd time l '1 is ovi t esta sh ant co itions itabl or the k cis n tb nce This consis nt with travel i 2.8 419 n diti .t TH POWER s ifie/in the tnote l -[M, id ch ged om t neric ue p vided NUREG-31 l to pla -s ic va e of a % RT. This ange i l_ cept le ause specif1 a in forpeo[er I agr nt ith c ent ope in rocedu ga )p cisi heat alance, urr t TS not' spec a power ( level for t s measu nt. imEgr3A-9 % l 05 08 Not used. 05 09 LG The requirements for inspecting and cleaning the feedwater flow venturi would be moved to licensee controlled documents. These details are not contained in NUREG 1431. This is an l example of moving unnecessary detailed information from the TS I and is acceptable. 05 10 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). i i 05 11 A Not applicable to WCGS. See Conversion Comparison Table f (Enclosure 3B). we _ -INSERT 3A -3). { CP 3.2 -oo i \\ [ Q2. A WCGS-Description of Changes to CTS 3N.2 9 2158 7

INSERT 3A-9a 0 3.4.1-2 5-07 A CTS 4.2.5.4 is revi:ed to reflect the current plant practices for r l performing this surveillance requirement. CTS SR 4.2.5.1 states that the provisions of Specification 4.0.4 are not applicable to RCS l flow rate. Maintaining the ISTS note allows performance of the SR in MODE 1 when the plant conditions support performance of the SR. The 24 hour interval in the ISTS is revised to a 7 day interval based on the plant prerequisites and previous operating experience l in performing the precision heat balance. Current plant practice is to perform this surveillance as soon as practical after the plant reaches 100% RTP after a refueling outage. Calibration of the instrumentation used for determining steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi Ap in the calorimetric calculations typically takes three shif ts (approximately 36 hours). Following instrument calibrations, data collection for the precision heat balance requires approximately two hours. Once the data has been collected, approximately 12 hours in necessary to complete the calculations associated with the precision heat balance. This changes is considered an administrative change since the change does not affect the operating limits or manner in which the plant is operated. l

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.2 Page 7 of 8 TECH SPEC CHANGE APPLICABILITY llUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY %-11 A_ INSERT 38 ~1_}----{ uC. 3.1-00 8 1 j 04-10 The al or t requir t to set t ower jler 4 A _ ti W *JA pe,,s4 X.a4 \\ Igef8-Ra Neutro ux-set nt dur powe educt k 032 -(. } ired QP TIONS Id xt o 72 rs idE Wolf Cr k. thyl-f.lsed. M i 05-01 The designation of how instrument uncertainties are treated Yes Yes Yes Yes LG (nominal in the analysis, or in the development of the CTS limit) is moved to the Bases. 05-02 The CPSES specific requirement to verify that the total RCS No Yes No No LS-7 flow is within limits using the plant computer or elbow tap output voltage on a monthly basis is deleted. 05-03 f The requirement __ m. Vos No-Herah cT5 Yes vee.e4.-Grt A 4Ts ves >5.- s1.t e cts LG

per L"..~.JA me L 4. _.a to normalize the 8CS looph rate 'mJk star.s. is

P...,~!; ac moved to the Bases for the surveillance requirements for tM "" 'k - ! =

r+ - t-fp f = tion cp 3,g.oo,1 3 1 w-ITS W .3.+.I 05-04 'Consis It h1~~ try t eler -105 y exp cit No - Requirement Yes Yes Yes LG r r ts th the R flow asur thr the se q not in CTS., [caIilprafian a ecisi heat ance me uremen* and th the \\ g Acth y wd -4hc 4^im 11 rument ion us in the forma e of. cal imetr' g gg,g.trubtation used in the prftirmana.- Q3,4 g~ g 1 l low me urement calib ted wi nas ift time g g gg g,c, pgu mem%,4 3 l gi of perf raing t measur nt is v_ed.___ the ases. rnw -to g Bas,es,,y 05-05 The Wolf Creek required actions would be modified to move No No Yes No LG details regardino identification of the cause for low RCS flow rate &.. '.. Tases. 05-06 The time t-redi . sower to less than 52 RTP would be Yes Yes Yes Yes LS-8 revised from + rhin 4 hours to within the next 6 hours. cti 3 CN C 05 -A. l l cts 4.2.S.4 is rev tied % reGect 4be. c.urvest+T ard 3.4-51. g a 3.4,s.1 { (pucticas & performin3 +% swvewmm reprementy WCGS-Conversion Comparison Table-CTS 3M.2 S/158 7

INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 i l TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF 26 incorporated 3.4-32 Approved by NRC. TSTF-27 Revh incorporated 3.4 33 M Q Q (Q 3.4/2.-Il 1 TSTF-28 Incorporated 3.4-22 Approved by NRC. l TSTF-54, Rev. I Incorporated NA Carp h MM'718.J.+#9l TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not incorporated Minor change that is adequately addressed in the Bases. 1 TSTF-87, Rev.h incorporated 3.4-31 (dpipd Q /rIL3.4,24. ] TSTF-93 @ Incorporated 3.4-17 hpred6)[RQ /QJ.4. 94} TSTF-94 Not incorporated NA Retained current TS. [7M 3.4 00 s- ( [TSJWI0fffInbrioratK _y M ~lQ 3.4.b 1 \\ \\ TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply. TSTF-113. RevhIIncorporated 3.4-39 q 3,4 ll-3] TSTF-114 Incorporated NA Approved by NRC. TSTF-116. Rev[' incorporated 3.4-36 { 4 3.4. 83 - 2 ] (AhevveAkgkNCD/72.3.6 dot f TSTF-136 Incorporated NA TSTF-137 Incorporated NA [A pprevsdL. 63 AMQ/7A 3.#- So f l TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6 TSTF-151 h Incorporated NA /T2 J.#409J TSTF-153 Incorporated 3.4-01 byAJA]/743.f-es? / TSTF-162 Incorporated NA {Appev vest by NAC]/73f.J.4*-20 6.} 'SOG4t3tef* D incorporated 3.4-4 % S )2. See also Cns 3.4-18 and 3.4-20.19 84IL* M (M *T5fFa$ incorporated 3.4-35 [4 3.4 nl.2 1 hWOG.69"Rawr0 Incorporated 3.4-10 DCPP onih5 ;d h NR".3N3 Mf I f ( WOb87,'Rev) Incorporated 3.4-47 [ 4 3. 4.11-4 } Q f 4 -2.} M-]2E Tncorporate ( + \\Q3.+.l').-Il Neh dpprerce.A. by 4 E A (W-tbeiimu..m.am.4. 5/15/97 -. = -.

l CHANGE l NUMBER JUSTIFICATION purposes (per Bases). This allowance is properly ) presented as an SR Note. A properly placed exreption (i.e., an SR Noted exception) would not allow the SR to be considered to be met until the appropriate conditions were available for it to be performed without entering the actions. The Note to these SRs would allow startup in i Mode 3 if the SR had not been performed during the required frequency. but would limit the exception to prior to entering Mode 2. The change is consistent with traveler @BlbfT5fF -@ q.3,4.ii-2. J 3.4 36 SR 3.4.13.1 and LC0 3.4.15 are revised per traveler TSiF-116. The note addresses the concern that an RCS water inventory balance connot be meaningfully performed unless j the unit is operating at or near steady state conditions. The note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours after re establishing steady state conditions. 3.4 37 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). j- (P.5.4.l-1 l 3.4 38 ( --P'Consi s nt T 105 the d ails o the met -e whi t CS f rat are v rified re mov fr of *PP 'Q* .1. to t Bases Movin this i formati t t-l U SeeConV**

N
Bas, al s the se of p cisio eat b nc (Endeswre. El 1

ps. .d other ccepta e met s in o er fo th verifi ion and s cons tent w ht N EGd431 p ilosop of movi clari ng inf a ' n pnd Qeser ive detai s out the T to e Bases. 3.4 39 The shutdown requirements of ITS 3.4.11 would require the plant to reduce T., to <500 F within 12 hours. rather than H0DE 4. to address the concern of entering [LTOP] LC0 = 3'"> 3.4.12 Applicability with inoperable PORVs., For ( e s.u-c.] consistency, the shutdown requirements of ITS 3.4.16 are also revised to allow 12 hours to reduce T., to <500 F. This change is consistent with TSTF 113. sTF 3.4 40 F Consi nt w trav r .t Not o SR .4.1. {Q3.4.1-21 wo be ified ' pro e ad iona time t perf a mg S pr sion f ra measu nt. he ti all ed ' wou te ch ed fr 24 ho s to 7 days. his an is ceptabl cau other dicati n of f1 is i avail e (SR .4.1.3. CS tot flow ter an d ional me no ly wou be r uir to stabbh WCGS-Differencesfrom NUREG-1431 - ITS 3.4 7 S/1S/97

_.__..m__._-___.-- INSERT 6A-7b Q 3.4.1-2 3.4-40 ITS SR 3.4.1.4 is revised to reflect the current plant practices for performing this surveillance requirement. CTS SR 4.2.5.1 states that the provisions of Specification 4.0.4 are not applicable to RCS flow rate. Maintaining the ITS note allows performance of the SR in MODE 1 when the plant conditions support performance of'the SR. The 24 l hour interval in the ISTS is revised to a 7 day interval based on the plant prerequisites and previous operating experience in performing the precision heat balance. Current plant practice is.to perform L this surveillance as soon as practical after the plant reaches 100%- RTP after a refueling outage. Calibration of the instrumentation -used for determining steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi Ap in the calorimetric l calculations typically takes three shifts (approximately 36 hours). Following instrument calibrations, data collection for the precision heat balance requires approximately two hours. Once the data has been collected approximately 12 hours in necessary to complete the calculations associated with the precision heat balance. Modify the note is consistent with current plant practices and the intent of the i CTS. I l j l l l I I 1 I -e:- ~ - e r y m m m -ewy

. ~. s CHANGE NUMBER JUSTIFICATION h 3. 4. ) - 2_ ortMrec [p~1antc ti 's abl on hea alance. Sin p hi ara er s no norma y chan c pnif anti and fl neter can be in t j in im. ere ~ no d to form t SR wit the 7da-l/p 4 ur iod cifi in N G 1431

v. 1.

T od pr ides ffici time t establis steady.cate t plant rmo rauli onditio and obt ' equi ' rium xe I dditio the TH L POWE pecif in te w d be c nged fr the gene c val in br kets (90 TP) t 95 % RTP This c ge is cept e se i specif s a power ev6 i tter r wit curren operatin rocedur for formi a pr isio3 r leveT for / hea alance. urrent J do no speci a l [ s measur nt. / J 1 3.4 41 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). 3.4-42 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 2 l 3.4 43 A new Condition is added to LCO 3.4.1 to reflect the current licensing basis of Wolf Creek for RCS flow rate. License Amendment 61 approvst revisions to incorporate the provisions of the RCS flow TS entitled "RCS FLOW RATE AND 4 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR" into the "DNB PARAMETERS" specification. These changes were made to support the use of VANTAGE SH fuel with the Intermediate Flow Mixer grid feature. This amendment also approved operation at an increased power level. ,503,4,g.2 1 s s c-2 3.4 44 Not li.ab jleo WCGSr -Tee f.onvffsiop-C-cifi'aniscii Tabte] p nel urea 5B). NSERT 0A-Sb l4 3.4.12,7. 3.4 45 ITS 3. 2 has n re sed move he Not for equ ed ) i Acti B.1 r ardin CP p swap operat' ns t licabi ty Not for a umulat isol ion t LCO. as dis ssed i ravel WOG- . Rev.f. P1 nt-cif

ti 11owan s for ceedin heLCJ'sn r f [E S]

,palEpscap e of jectin. nto th e i corpo at N 3.4183. These,e'RCS j as disc sed in Not de il s' ua ons wher except ns to tJxILC0 a/rpermi ted nd a e i a ropriat y annotated undpf' the L (DSEd.T GA-pa.-) ~ -- 3.4 46 Consistent with current TS 3/4.1.1.4. " Minimum Temperature for Criticality." ITS LC0 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current TS wording is acceptable since valid WCGS-DifferencesfromNUREG-1431-1TS3.4 8 S/1S/97

=. - CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 6 01 o SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER DESCRIPTION ocpf DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY r %;.wb% %e' cts sd 4.2.3.3,the detah on 4.a. Euh which-4 t.4.+msk. +he. EESfte,w < ate io ver LA sk_ veMjvwws STs. ssJa._ detat on/he't by Yes Y* No MMo X {Q3 4.t-Ij 3.4 38' Consts ~ ~ with48TF7.t whi the fRflow r e ar erif' arfeo f SR 3 f.1.4 to _ Ba f-3.4-39 The shutdown requirements of ITS 3.4.11 would require Yes Yes Yes Yes the plant reduce T, to <500T within 12 hours. rather than MODE 4. to address the concern of entering [LT0p] LCO 3.4.12 Applicability with inoperable PORVs. For corisistency. the shutdown requirements of ITS 3.4.16 are also revised to all 12 hours to reduce T, to <500T. This change is consistent with TSTF-113. N4o 5re.] 3.4-40 kNo to 3.4.1.4 would modifi do ya No - See CN No - See CN Yes ic reac+ # power nd to ovide itidial 34 _ 3.4-34 e.Q[q3 c4 3.4 6E.L_ pla s 1sion ow rat rrs sg 3.4.,).4 i3 revtSed b otfleck the_ cs

4. t-2.

rfors/RCSp I me t g%,.4,,gAna M g; Qa rement. / f pl>d p,,,chca s,4, gpq ::ha ~ y 1 w 3.4-41 LCO 3.4.1 is revised to reference Tables 3.4.1-1 and Yes - Allowance No No No 3.4.1-2 for RCS total flow rate limits for DCPP Un.ts added per I and 2 respectively. Amendment 60/59. 3.4-42 An exception to SR 3.4.14.1 frequency to leak test Yes - Specific No No No PIVs 8802A, 8802B and 8703 has been added. This to DCPP change is consistent with the DCPP current TS. 3.4-43 A new Condition is added to LCO 3.4.1 to reflect the No No Yes No current TS of Wolf Creek for RCS Flow Rate. 3.4-44 Steam generator levels for MODES 3. 4 and 5 are No No JD-Yes 43G '2-PM"3 _ specified to ensure SG tubes are covered. The g 0;af36@)currer.t TS did not ensure tube coverage. N i S/lS)97 WCGS-Conversion Conperison Table-ITS3.4 i s

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.5-2 APPLICABILITY: CA, WC REQUEST: ITS SR 3.4.5.2 (also SR 3.4.6.2 and SR 3.4.7.2) (Callaway) Change 1-15M Comment: The sections of the ITS use the phrase "or equivalent" yet the term is not explained in the change or in tha ITS Bases. According to the information provided narrow range level is used at the higher temperatures (Modes 3 and 4) and wide range level is used at the lower temperatures (Mode 5). If "or equivalent" means using the wide range at higher temperatures and the narrow range at lower temperatures are the levels specified applicable at the different temperatures? If not, what are the equivalent levels to the values specified in the ITS and how were they determined? FLOG RESPONSE: (original) At Callaway, the top of the highest steam generator (SG) tube is 3~ i inches above the tube sheet. The wide range instrumentation provides level inoication from 7 inches above the tube sheet (0% indication) to the moisture separators (a range of 559 inches). The narrow range instrumentation provides level indication between 438 and 566 inches above the tube sheet for its 0-100% indication (the use of a common upper tap results in 100% level indication on both wide range and narrow range nominally being at the same 566 inches above the tube sheet). A calculation was performed to correlate the top of the highest tube to the wide range scale for MODE 5 conditions (the wide range instrumentation is calibrated for cold conditions), with margins added in for instrument loop errors and readability, resulting in the specified 67% wide range level. A minor error in the calculation was corrected, resulting in the specified 66% wide range level value cited in the attached pages for Callaway. Since the zero reference for the narrow range level instrumentation is nominally 96 inches above the top of the highest tube, the 4% value specified for MODES 3 and 4 was chosen since it is used throughout the EOPs for heat sink indication and is familiar to the operators. In the main control room there is one Class 1E wide range level indicator per SG and there are four narrow range level indicators per SG, of which three per SG are Class 1E. The "or equivalent" phrase would allow the use of wide range levelinstrumentation in MODES 3 and 4 in the unlikely event all narrow range level instrumentation were unavailable for a required SG; in MODE 3 this unlikely scenario would result in ITS LCO 3.3. > non-compliance and would invoke Required Action (s) under PAM Instrumentation. Conversely, the "or equivalent" phrase would allow the use of narrow range level instrumentation in MODE 5 if the one wide range levelindicator per SG were unavailable. This flexibility is similar to the approach under which Vogtle was licensed wherein their MODES 3-5 RCS specifications required SG water level to be above the highest point of the SG U-tubes. We are specifying water levels that ensure the same, yet allow the use of all available instrumentation. Before the "or equivalent" instrumentation were used in a given MODE, process measurement effects on the alternate instrument's calibrated span would be considered. Due to the unlikely event of either scenario presenting operational limitations, given the reduced RCS loop l l l l

. -. -.. ~ -. - -. -, requir;;mants in MODES 3-5 and the instrumentation redundancy, we do not see the need for a pre-determined correlation between the wide range l and narrow range level indications; however, we reserve the right to exercise that option should the need arise. Wolf Creek reviewed this particular comment for applicability to Wolf j Creek and concurs with the use of the phrase "or equivalent" in the ITS l and ITS Bases. Wolf Creek believes that it is appropriate to change their plant-specific value to 6% narrow range (including uncertainties) since it is used throughout the Emergency Operating Procedures (EMGs), it has operator awareness because of the EMG familiarity, and ensures an SG i water level approximately 100 inches 'above the top of the highest SG L tube. Wolf Creek has done a review of the drawings and design l documents and has determined that for MODE 5 conditions (the wide ) range instrumentation is calibrated for cold conditions),66% wide range l level corresponds to the top of the highest tube, with margins added in for instrument loop errors and readability. The need for flexibility to use either narrow range or wide range indication is most evident when placing the SGs in wet layup conditions. The narrow range instruments are "jumpered" to indicate a constant 50% level. This precludes a feedwater isolation signal at approximately 78%. The operators use SG wide range indication to maintain and monitor SG level. Additionally, the narrow range instruments are calibrated for normal operating pressure and temperature i conditions while the wide range instruments are calibrated for shutdown conditions. I The Callaway and Wolf Creek ITS Bases have been modified to explain l the "or equivalent" phrase. i FLOG RESPONSE: (revised) ITS 3.4.5 and ITS 3.4.6 have been revised to refer only to the narrow range SG level indicators (4% for Callaway, 6% for WCGS). ITS 3.4.7 has been revised to refer only to the wide range SG level indicators (66%). The "or equivalent" phrase has been deleted. The Bases for ITS 3.4.7 have been revised to reflect that any narrow range level indication above 4% for Callaway (6% for WCGS) would ensure the top of the SG i tubes are covered. This is a valid statement since MODE 5 temperature conditions will not induce a 100 inch level measurement error on the i narrow range. Since this is a revised response, all of the required change pages from Reference 5 of the cover letter are attached. . ATTACHED PAGES: Attachment No.10, CTS 3/4.4 - ITS 3.4 Encl. 2 4-2, 4-4, 4-5 L Encl. 3A 3 Encl. 5A 3.4-11,3.4-13,3.4-15,3.4-17 Encl. SB B 3.4-26, B 3.4-32, B 3.4-33, B 3.4-31, B 3.4-35, B 3.4-36 i l

REACTOR COOLANT SYSTEM 1 i HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least ^f.n: ' 5: @ actor coolant loops listed below shall be 1-14 LS-22 i OPERABLE and at least two of these reactor coolant loops shall be in operation "when the Rod Control System 6s capable of rod withdrawal and at least one 1 reactor coolant loon in onoration when the Rod Control System is not capab!s} 4 Lof rod withdrawalft"J

a. Reactor Coolant Loop ^. :nd 5 rr: ' ': f :'rm ;;xx^:: nd

~ 1-01-LG l. -^ ^ ' '-' : : : 'l ^ -"-, 4

b. Reactor Coolant Loop h;:nn.:::nd

-- - r :':..:; -.;,

c. Reactor Coolant Loop C :nd 5 xx- :f 'rm ;:nn - :::d

-- "- x:'z:; r;.: d

d. Reactor Coolant Loop D :nd b :- : : :f :'r ;x:._:: d

- " x :' n : ; n. APPLICABILITY: MODE 3.2 1-04-M ACTION:

a. With ':x " n 5: 11;: @equired reactor coolant loops &E14ABLE

.. 11-19-M [Inopersnesarestore the required loops to OPERABLE status within 72 hours or be ~ ~ " " in HOT SHUTDOWN within the next 12 hours.

b. With only one reactor coolant loop in operation,-eestese el-least 1-14-LS-221 5 ' :;: t: :;:-N :t:- 72 5 : er '"'-

5er rn N

{ rod witherawal, within 1 hour restore two loops to operation or plac]e .,,___... _ '- -'nd the rod control system capable of j the rod control system in a condition incapable of rod withdrawal. g

c. Withffour RCS loops inoperable olino reactor coolant loop in operation, 3-04-LS-29 F immediately place the rod control system in a condition incapable of rod]

Iwlthdrawalfluspend all operations involving a reduction in boron concentration of $1-19-M 7 the Reactor Coolant System and M ' '"; initiate corrective action to retum the " " ~ ' ' " ' ~ ~ " ' esquwed one reactor coolant loop to OPERABLE status and operation. SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by venfying correct breaker ali noments and indicated power availabiiny.

61. p pi,@ -j cm.15.4.5 }

} Q3.4.5-2,diL4.G-3 l 4.g.1.z.z Ine requirea steam generators shall be determined OPERABLE by ing secondary side wide-InarroWirange water level to be greater than or equal to "1 15-M at least once per 12 hours. ~- I 4.4.1.2.3 ^* ':r' d Jhe required} reactor coolant loops shall be verified in operation 1;14-LS-22'

nd :: ' n; n "- x :'z: at least once per 12 hours.

M1 01-I.E, .~umm

All reactor coolant pumps may be_f: n;:;Mremoved from operation]

V1-16A" for up to 1 hour {per 8 hour periodbrovided: (1) no operations are permitted

gg.;

that would cause anution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

- E:: ?:- " T-
i:_;^J:-- ?; r:'r:n ? * ^ '

l-04-M J "No RCP shall be started with any RCS cold leg temperature 5368'F unless the l 105-M secondary side water temperature of each steam generator is s 50'F above each ~ [ of the RCS cold leg temperatures. j WOLF CREEK-UNIT 1 3/4 4-2 Ma k-up ofCTS3N.4 S/2S/97

~ REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4 l 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pump. if not in operation. shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.3.2 The: required steam generator (s) shall be determined OPERABLE by l ' 'ng secondary side Cnarrownge water level to be greater than or equal 1-15-M - 'v to at least once per 12 hours 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation =d . '-' -; :--'- :::'r.: at least once per 12 hours. 1-01-LG L o 3 A.5-2. 6 j 4 S 4 5% (fase.s-2.1 WOLF CREEK - UNIT 1 3/4 4-4 Mark-up of CTS 3N.4 $/15/97

a 4 REACTOR COOL. ANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 4 I 3.4.1.4.1 At t one residual heat removal (RHR) loop shall be OPERABLE and i?b88-IK2*8 in operation and either: j q :!!,. 4.5 j husma

a. One additional RHR loop shall be OPE (g,g, gg Q 3.4.s-2.,

i

b. The secondary side w '

vel of at leasuwo..... v -.wi. @ 3.Md i 15-MT" ??gw.u.ac.sM shall be greater than of 0 :~WWange. f j APPLICABILITY: MODE 5 with reactor coolant loops filled ##. ACTION:

a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to retum the inoperable RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.

}

b. With[requ6 red RHR loops inoperable o3no RHR loop in operation, 3111M]4 suspena an operations invohnng a reduchon in boron concentration of

- the Reactor Coolant System and ;,,,,re' ^

initiate corrective action to retum the required RHR loop k(OPERAE LE status and)peration.

f SURVEILLANCE REQUIREMENTS N i-4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4 4.4.1.4.1.2 At least or,e RHR loop shall be determined to be in operation-and Ilbl# M'^^t;:- :': ::":.::tleast once per 12 hours. [ (NEW) Verify correct breaker alignment and indicated power are available 7*1kil-MC. i to the required RHR pump that is not in operation at least once per 7 days.

One RHR loop may be inoperable for up to 2 hours for surveillance testing prwided the other RHR loop is OPERABLE and in operation.

4 1 ' ##A reactor coolant pump shall not be started (with any RCS cold leg temperature) % 1-05 M I i unless the secondary water temperature of eacn steam generator is less an O'F above each of the Reactor Coolant System cold leg temperatures. l

The RHR oumo may be f:::.;; :dromoved from operation #or up to I 41Mk]

1 hour [gier 8 hour periodjprovided: (1) no operations are permitted that would cause "M{d ji l dilution DT me Reactor Coolant System boron concentration, and (2) core outlet ts:. V temperature is maintained at least 10*F below saturation temperature.

- All RHR loops may be removed from operation during planned heatup to MODE 4

'l88-lE-27 when at least orwn RCS loop is in operation. 3j WOLF CREEK-UNIT 1 3/4 4-5 Mark-up ofCTS 3M.4 S/l587

__m CHANGE NLPEER EHC DESCRIPTION 1-11 M This change adds a new surveillance for verification of breaker alignment and power availability to the required pump not in operation. This change is in conformance with NUREG 1431 Rev. 1. 1 12 M The Actions are changed to separate the required actions for only one required RHR loop OPERABLE and no required RIR loops OPERABLE.- These revised Actions are consistent with the Actions which are required under this LCO in NUREG 1431 Rev. 1, and are more conservative than current required actions. 1 13 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 1 14 LS 22 The L and ion are r sed ' req re yt 1 s l OP LE th t loops oper ion nt r co 1 yst s ca le of wit rawal o 1 in o ation nt od co 1 sy em i not pa eo wit rawal This c nge i onsi ent th 1431) Rev. INssCT 3A-3 a 1 15 H A steam generator (SG) level corresponding to 10% of the i wide range does not cover all of the SG tubes. To qualify as a valid heat sink, the tubes must be covered. This is a more restrictive change. [,] g gy ~ 1 16 A Consistent with the intent of traveler TSTF-153. this change revises the note that permits up to 1 hour "deenergization" of RCP/RHR pumps. The revised wording clarifies the intent of the note to allow the pumps to be " removed from operation" instead of "deenergizeu", thus permitting other means of removing the pumps from service. With this change the pumps are not reauired to be deenergized to use the note (e.g. the pumps may be i nlated, etc.). The change is considered to be ddRinistrative because from the standpoint of providing an exception to the LC0 requirements (to maintain the operability and operation of the pumps), the revised wording is equivalent. 1-17 LG Not applicable to WCGS. See conversion Comparison Table (Enclosure 3B). WCGS-Description of Changes to CTS 3N.4 3 S/15/97

~ INSERT 3A-3a 0 3.4.5-1 The LCO and ACTION b of Specification 3.4.1.2. " Reactor Coolant System, Hot Standby," would be revised to require that two reactor coolant loops be OPERABLE. Loop operation requirements would also be revised to be contingent on Rod Control System status. The requirement to have a third OPERABLE res tor coolant loop would be deleted, consistent with NUREG-1431. This is acceptable because the MODE 3 decay heat removal requirements are sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay heat. A second RCS loop ensures redundant capability for decay heat removal. When the Rod Control System is capable of rod withdrawal, two loops must be in operatien to ensure accident analysis assumptions are satisfied. When rod withdrawal is precluded, only one loop is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident analyses which assume only two RCS loops in operation include the Uncontrolled RCCA Bank Withdrawal from Subcritical and the hot zero power RCCA ejection events. The initial conditions and analysis assumptions for these events will be unchanged since two looot sust still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal. These reactivity transients rely on the Nuclear Instrumentation System's high flux trips for event termination which occurs very rapidly (on the order of seconds). There would be no oenefit of having a third RCS loop OPERABLE for these transients since by the time the loop could be brought into operation, the event would be over for all practical purposes. INSERT 3A-3b 0 3.4.5-2 0 3.4.5-3 Six percent of the narrow range span is specified at the higher temperatures of MODES 3 and 4 whereas 66% of the wide range span is specified for MODE 5. Both values ensure SG tubes are covered. The Emergency Operating Procedures cite the 6% narrow range level to ensure heat sink adequacy. l l l l

l l RCS Loops-H00E 3 j 3.4.5 l i SURVEILLANCE REQUIREMENTS ( SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation. 12 hours 4 l kb d" SR 3.4.5.2 Verify steam generator secondary side water 12 hours ~ levels are a W4@>gprojpgogffor required gg; RCS loops. 3.4- -- -- - % (q s.4.s.t 1 $$.4..so l l z i SR 3.4.5.3 Verify correct breaker alignment and indicated 7 days l power are available to the required pump that is not in operation. l l l I l I i l l WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 11 S/15/97 l

RCS Loops MODE 4 3.4.6 ACTIONS (continued) COWITION REQUIRED ACTION COMPLETION TIME B-0 4 required = leep B-1 "; ir. O C 5. 24 hours $$i458Eff irgrdle. l M l Tw r;;; ired "CC [ leep; ir g r 212. i GB. Required "CC er = GB.1 Suspend all operations Inmediately N9]F02d i loops inoperable. involving a reduction of RCS boron DB concentration. l No RCS or RIR loop in M operation. GB.2 Initiate action to Innediately restore one loop to OPERABLE status and operation. I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. 12 hours l SR 3.4.6.2 Verify SG secondary side water levels are a W 12 hours ggy l hh for required RCS loops. 3 4,44 l W$ hhk N3- _93 Nobs.21 (continued) WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 13 S/1SM7 r

RCS Loops HODE5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-H00E 5 Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either: os.4.5 i a. One additional RfR loo shall be OPERABLE: or @s.4.s3 under @ b. The secondary si water leveFor at least two steam generators (SGs) shall be :t 3 .4 44-) ........................... NOTES ggxs:p 1. The RIE pump of the loop in operation may be de cr.crezed

$324!00M reWoveid f@m: operation;for s I hour per 8 hour period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and b. Core outlet temperature is maintained at least 10*F below saturation temperature. 2. One required RHR loop may be inoperable for up to 2 hours for surveillance testing provided that the other RHR loop is OPERABLE and in operation. 3. No reactor coolant ptmp shall be started with one-ee-mere NNMN any;RCS cold leg temperatures s-4752F 368'F unless the

gypsy secondary side water temperature of each SG is s 50*F above each of the RCS cold leg temperatures.

dN8I N 4. All RHR loops may be removed from operation during planned heatup to H00E 4 when at least one RCS loop is in operation. APPLICABILITY: MODE 5 with RCS loops filled. WCGS-Mark-up ofNUREG-1431 -ITS 3.4 ' 3.4 15 S/15/97

l 1 RCS Loops HODES. Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY bk"N hIisI, #^ SR 3.4.7.2 Verify SG secondary side water level is 2-W 12 hours W57,bjin required SGs. $M /MhN

4"#4' u

. q,,,,,3.y SR 3.4.7.3 Verify correct breaker alignment and indicated 7 days power are available to the required RHR pump i that is not in operation. i l i 1 l l WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 17 5/15/97

RCS Loops - MODE 3 B 3.4.5 BASES SURVEILl/NCE SR 3.4.5.1 (continued) REQUIREME!.TS considering other indications and alarms available to the operator in the control room to monitor RCS loop performance. SR 3.4.5.2 k 3.d SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is g varif by ensuring that the secondary side narrow range water level is a for required RCS loops. 4 If the SG secondary side narrow range ~ water level is < , the tubes shy become uncovered and the associated 6 'I. loop may not be capable of provid' the heat sink for removal of the decay heat. The 12 hour Frequenc is considered adequate in view of \\os.4.5-2.} other indications available in t control room to alert the operator os.45-31 to a loss of SG level, f --- ,i p e i u e k o s. 4-ev at a i SR 3.4.5.3 ,f r n a t in w av ^ A meu.ss\\ ~ Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed. to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required RCPs. REFERENCES Nene: IJ USAR iSection 15.4.6 2; IRC,W3W.5Reckley to~NOCarnsEdated November.22,11993: Wolf Cese_k2SeneratingiStation -iPositive;Rea.ctivity Addition;l. Technical Specification _ Bases; Change." WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 26 S/1/SM7 \\

RCS Loops-H0DE 5, Loops Filled B 3.4.7 B 3.4 REACTOR C00LAKT SYSTEM (RCS) l B 3.4.7 RCS Loops-MODE 5, Loops Filled l ) BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer cEthis heat either to the steam generator (SG) secondary side coolant t{aPnBtiFaTifftg18tTdhWf'E or the component cooling water { via the residual heat removal (RlR) heat exchangers. While the principal means for decay heat removal is via the RIR System, the SGs are specified as a backup means for redundancy. Even though ' the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor ) coolant is to act as a carrier for soluble neutron poison, boric acid. In MODE 5 with RCS loops filled, the reactor coolant is j circulated by means of two RR loops connected to the RCS, each i loop containing an RHR heat exchanger, an RHR ptmp, and appropriate flow and temperature instrumentation for control, protection, and indication. One R m pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratificationJtiis.pt; sufficient"for~the boron ~ dilution analysis discussed;.be]og The ntaber of loops in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at lesst one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal. The LCO provides for redundant paths of decay heat removal capability. The first path can be an RHR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop o maintaining two SGs with secondary side water levels above to provide an alternate method for decay heat removal via~ _ tura1l circulation. (Reff 3). ~ Q 3. 4. O"4 O * % *- U Q 3.4 5-3, (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 32 5/1/S87

h RCS Loops-MODE 4 B 3.4.6 l BASES 53 d'd SUR LANC SR 3.4.6.2 6 4 j'j (continued) SR 3.4.6.2 requires verifica on of SG OPERABILITY. SG OPERABILITY is verified by en ng that the seco ry side narrow range water level is a J, M rts .,ptgij. f the SG secondary side narrow range water level is < the tubes ( may become uncovered and the associated loop may t be capable of providing the heat sink necessary for removal of decay heat. The 12 hour Frequency is considered adequate in view of other indications available in the control room to alert the operator! to the loss of SG 1evel. SR 3.4.6.3 + e. n Verification that th required ptmp is OPERABLE ensures that an N14N additional RCS or RfR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment l and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative l controls available and has been shown to be acceptable by operating experience. l l REFERENCES Nene-1~. t W E SectionJ 5 4f.6 i l 2; 16tC11etterf(W6Reckley to NECarns); dated November:22; 1993j[1'llolfl Creek (GeneratingjStationd@sitive Reactivity L AdditioniTTechnical15peciffcation Bases Change ~.~" l l t i WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 31 S/1/58 7

. _. _. _ _ _ _ _._ ~. RCS Loops-H0DE 5, Loops Filled B 3.4.7 4 4 BASES (continued) 4 APPLICABLE In H00E 5. RCS circulation is considered in the SAFETY ANALYSES determination of the time available for mitigation of the accidental boron dilution event. The PJP leep previi this circul; tion. J TheJoperitiKoEone RCP2in900ES 3k4Tandl5 provides:'adeguite fjewit_o; ensure,.mixingEgeventJs_tratificattogid and goduce 4 geltle!!IidinctM13gchinges3ringus~ boron concegrat16n olikktfensDiithino2eactor2cooligErdoplMapersDonMytthit i HllDESWCorEboronfdi18ttonslaustibe3erminhtedland31 Mon 3 scorencp!! ate 3tane; boron d1DtioniadaMs1MMEDESMe l carit;7arythitutginsolumelassociatedMth;naimg xinsttime reeMoCoitoJjnt IdopMpperatibgCTRef21); RCS Loops-HODE 5 (Loops Filled) Mvc ban iintificd in the h"'O relicy Stetc.;;nt a impertent centributer; to ri;k reduction sati sfi.esf eri teYibrF4"of'10 ~ CFR' 50.36(c) (2) (iiT.~ 45I i w r,~s ~) LCO The purpose of this LCO is to require that a least one of the ps.45 2. l RHR loops be OPERABLE and in operation with additional 43.4.5.3, loop OPERABLE or two SGs with secondary side ater level x u-. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under these conditions. An additional RHR loop is required to be OPERABLE to j meet single failure considerations. However, if the standby RHR loo) is not OPERABLE, an a able alternate method is two SGs 4 4 wit 1 their secondary side range water levels 2 I 4,34.s-2. ~ } Should the operating RHR loop

11. the SGs could us to qs.4.s.3, V

h

8 remove the decay heat via1 natural circulation.

a e att GG% C Note 1 permits all RHR pumps to be removed from operation for e a i i f & caergiced s 1 hour per 8 hour period. The purpose of the Note M J-is to permit tests that are requiredito be performedyithoutiflow oriptuo noise. i;igrad to valiit; various oxidcat ;mly=; fny narrow nrv3 y;;;;;, ~ 0 2 ;; tg t;;;; p;7;;7;;;;g 3;7;;; th; ;;;7t;p t;3 ting is. vel thcuc* hon i preire;;; is tk veliitien Of red drep ti;;;;s during c;ld [ above. 6%JWI *N-conditism, kth with end with;ut flow. Thc n; fl;W tcst =y M .the sG-bbes, >m j pcrfor;;;d in "00: 2, 4, or S and rquirc; tMt tb pu;;;ps bc covered.. / stepp d for ; ;M rt period of ti x. Thc tt; pcr;;iits i crergicin; cf tk pu;p; in order to perfer;;. this tot end validet; th; n;u cd : =ly;is volucs. If cMn;cs arc mi t; the RCS th;t w;uld au x ; c M n;; t; the fi;w cb rectaristics of the 1 RCS. the input v;1ue; ;r,u;t bc rcv;litted by conducting th; tcst egeift-The 1 hour time period is adequate to perform the necessary testing, and operating experience has shown that boron i (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-33 S/1/SR7

1 RCS Loops-MODE 5. Loops Filled B 3.4.7 l BASES this LCO requires forced circula;MjastIhMmst{q In MODE 5 with RCS loops filled f APPLICABILITY' 1on or Ine reactor coolan: a remove decay heat from the core and to provide proper boron mixing. One loop of R}R provides sufficient circulation for these purposes. However,: one additional RHR loop is required to be OPERABLE, or the secondary side (1NEiEH'ar!ge water level of at 0 3 4.s-2. least two SGs is required to be a t 66 */. 4 'A. 4.C *.5 Operation in other MODES is covered by: LCO 3.4.4, "RCS Loops-MODES 1 and 2": LCO 3.4.5, "RCS Loops-MODE 3"; l LCO 3.4.6, "RCS Loops-MODE 4": l LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled": l LCO 3.9.5, " Residual Heat Removal (RtR) and Coolant Circulation-High Water Level" (MODE 6); and LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6). L l ACTIONS A.1 and A.2 Q A4.5-2. \\ ps.4.s 3 ,m If one RHR loop is rable and the required SGs have secondary W M w f-) side 5 tater levels 4 redundancy for heat removal is lost. Action must be initi immediately to restore a second RfR loop 4 to OPERABLE status or to restore the required SG secondary side l water levels. Either Required Action A.1 or Required Action A.2 l will restore redundant heat removal paths. The 1:enediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal, i j B.1 and B.2 If no RHR loop is in operation, except during conditions i l permitted by Notes 1 and;4, or if no loop is OPERABLE, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE I l status and operation must be initiated. Additiori3fEbbrit4d j. water ~irithy:6oncentration greater;than.;or(eiqualjto:;thelminimum i required; RMS.Ticoncentration: bat;,less; than;the;act.ua]Rbbron concentration ~shallfnot' be~ considered'a :reductionlinfboron concentration. (Refs 2). Topreventinadvertent:critjcality duringla boron dilution, forced circulation frouf atleast~one RCP r is required to provide proper mixing. ;-4 r;xrs; th; 27;17. tc critic;11ty ir, thi; t g; cf e g r;ti;r.. The immediate Completion Times reflect the importance of maintaining operation for heat removal. l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 35 S/VS/97

RCS Loops-H00E 5, Loops Filled B 3.4.7 BASES (continued) SURVEILLANCE SR 3.4.7.1 This SR reqm-verification every 12 hours that the required loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available j to the operator in the control room to monitor Rm loop performance. i 'ai.4 5 O tide. {a3.451 ( r -- Verifying that at least two SGs are OPERABLE by. Any norma rmf. ringtheirQu.so( f I* d tndag secondary side range water levels are 2 ( ensures an , y* 4 4 jgt a alternate decay removal method islavailab1_iX]!jatgral cifcGletion in the event that the second RHR loop is not (b * * *j OPERABLE.h If both RHR loops are OPERABLE, this Surveillance covered.. i not needec The 12 hour Frequency is considered adequate in view of other indications available in the control room to alert the_ L operator t the loss of_ SG 1evel. g3.45 2.] i ia y- $) - SR 3.4.7.3 1 'A Verification that a second RHR pump is OPERABLE ensures that an gg, l additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. l Verification is performed by verifying proper breaker al t We and oower available to RHR pump. If secondary side range water level is in at least two SGs, this Surve ance is not needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES None-1~ USAR,;Section 15.4 6 2. EC'. etter (W. Reckley, to N.;Carns)1 dated _ November.22, l 1993:'" Wolf Creek; Generating ~Stationj; Positive Reactivity Addition;[ Technical _ Specification _ Bases ~ Change." 3. EC"Information Notice 95 35,1* Degraded l Ability of SGs to Remove Decay Heat by Natural Circulation." i WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 36 S/1/38 7

l ADDITIONAL INFORMATION COVER SHEET i i ADDITIONAL INFORMATION NO: O 3.4.11-3 APPLICABILITY: CA, CP, DC, WC REQUEST: Change 4-05 LS 31 and Difference 3.4-39 Comment: TSTF-113 (presently Rev. 4) has not yet been approved by the NRC staff. l l FLOG RESPONSE: (original) TSTF-113 Rev. 4 revises the shutdown requirements of ITS j 3.4.11 to allow the plant to reduce T., to <500 F within 12 hours, rather i than MODE 4, to address the concern of entering LCO 3.4.12 Applicability I with one or more inoperable PORVs. The shutdown requirements of ITS 3.4.16 are also revised, for consistency, to allow 12 hours to reduce T,y i to < 500 F. ITS 3.4.11 Condition B and C Bases changes have been l made to the Callaway submittal to reflect Rev. 4 of the traveler; no l changes are required for any other plants'submittals. The FLOG continues to pursue the changes proposed by this traveler. FLOG RESPONSE: (supplement) Based on the present status of the generic change l' process for the STS, it appears that traveler TSTF-113 will not be l approved by the NRC in time to support the initial license amendments for l the FLOG plants. In order to facilitate the issuance of these initial license l amendments, an alternate approach has been developed by CPSES and DCPP which relies on the CTS, plant-specific information, and/or the NUREG but does not rely upon the traveler. This alternate approach is hereby provided as an interim submittal to allow issuance of the initial license amendments for CPSES and DCPP. The changes which rely upon the traveler can be processed in subsequent license amendments following approval of the traveler by the NRC. Callaway and WCGS continue to pursue the changes proposed by this traveler as a beyond scope change, primarily due to their high COMS/LTOP arming temperature of 368 F. Required Actions in STS 3.4.11 for one or both PORVs inoperable would bring the plant to MODE 4 (350*F). This could involve entering the COMS/LTOP LCO Applicability l (ITS 3.4.12) with no means to mitigate a cold overpressure event since RHR is just being aligned for shutdown cooling during this time frame. Callaway and WCGS propose to terminate the shutdown tracks in ITS l 3.4.11 at an RCS T,y < 500 F, rather than 350 F. At T,y < 500 F, i saturation pressure is 666 psig. The lowest MSSV lift pressure is 1185 psig with a +3/-1% setting tolerance. The SG atmospheric steam dumps (ASDs, WCGS refers to these valves as the SG atmospheric relief valves or ARVs) have a set pressure of 1125 psig. Operation in MODE 3 with RCS T,y < 500 F renders the offsite release of radioactivity in the event of an SGTR extremely unlikely given: 1) the difference between the initial SG pressure for this RCS T,y and the ASD/ARV set pressure; 2) the secondary side pressure increase would be driven by RCP and decay heat only since the plant is in MODE 3; and 3) operation of the steam dump to the condenser. The licensing basis SGTR analysis discussed in FSAR/USAR Section 15.6.3 uses worst case assumptions (e.g.,3636

MWt initial core power,18 MWt RCP heat, and 939 psia initial SG pressure for Callaway corresponding to 15% tube plugging) and ignores - the steam dump to condenser to intentionally drive the secondary side pressure above the ASD/ARV lift setting to evaluate offsite doses. When consideration is given to the extremely low probability of lifting the ASDs/ARVs after an actual SGTR with RCS T., < 500 F versus the real potential for enteririg COMS/LTOP LCO Applicability with no mitigation pathway, Callaway and WCGS maintain that the proposed change results in an enhancement to net plant safety. ATTACHED PAGES: None

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.12-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-49 Comment: WOG-100 has not yet become a TSTF. FLOG RESPONSE: (original) WOG-100 has been approved by the TSTF and is designated as TSTF-280. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-280 was modified from WOG-100 and these modifications have been incorporated into the ITS (added "or" to LCO list and SR 3.4.12.5 Note was deleted). The FLOG continues to pursue the changes proposed by this traveler. FLOG RESPONSE (supplement) TSTF-280 has not yet been approved by the NRC. This traveler is being withdrawn. In place of the traveler, the ITS is being revised to match the CTS and to incorporate some clarifying editorial changes as agreed upon in a conference call with the NRC staff on November 19,1998. ATTACHED PAGES: Attachment No.10, CTS 3/4.4 -ITS 3.4 Encl. 5A Traveler Status page Encl.6A 9

INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-26 Incorporated 3.4-32 Approved by NRC. TSTF-27. Revh incorporated 3.4-33 {Appr,M ggQQ 3.4.'2.-Il I TSTF-28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev. I Incorporated NA CAh_P "M rsa J. +.m9 l TSTF-60 incorporated 3.4 15 Approved by NRC. TSTF-61 Not incorporated Minor change that is adequately addressed in the Bases. TSTF-87, Rev.h Incorporated 3.4-31 (hodby UQ hW 3. X)84 J (Aphe <v./ 6[hRQ /QJ.4. 9-3] TSTF-93 Q Incorporated 3.4-17 TSTF-94 Not incorporated NA Retained current TS. [TM 3.4.co s-[ [TSTMTO V / /_ _ _InMoratR 7.A,336-l 4 3.4. b Ll TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does r.ut apply. TSTF-ll3, Revh incorporated 3.4-39 4 3,4, gg.3} TSTF-il4 incorporated NA Approved by NRC. TSTF-116. Revh' incorporated 3.4-36 16 4 83-M (APhrevcA. hMC] /pt.3.5co? / TSTF-136 Incorporated NA TSTF-137 incorporated NA @ppemd. by NMJ/7A 3.4-So9 I TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6.. TSTF-151 h incorporated NA /7R J. W l hewvc4 by MAC]/7MJ.f-dt39/ TSTF-IS3 Incorporated 3.4-01 (A[p$owd by NAC.)/TW.J V'- $0s-l TSTF-162 Incorporated NA IS4KT-WJ5ef"3 Incorporated M 3.4-4. % f ]2. See also Cns 3.4-18 and 3.4-20.lq s.4.a t_ ) (W4C90 ~75fF-2Ap Incorporated 3.4-35 [4 3.A,il -2.\\ IbOC WHs><D Incorporated 3.4-10 DCPP onlykprawd h MMl7EJ4-oo?7 ( WOU-87, Rav3 Incorporated 3.4-47 14 3.4.H-4 i ~ N3* Incorporated 3.4-40 Applicable to Callaway and (T*fs:-28ig Wolf Creek only. {@ 3.4.1-2-} hM2 incorporateWh --f" %.3.+. I'2. - I j Nd i w Lgrevt.A. by satu 4 _. _u. 5/15/97

CHANGE NUMBER JUSTIFICATION T,,, measurements are not obtainable for a non operating loop. 3.4 47 ITS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Condition B or E of LC0 3.4.11. However. Required Action A also directs closure of the block valve whn one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also i exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 in as much as the block valve cycling is exempted under Conditions A.,B. and E.fS nte _ ~ toA he4 1ocF4alue M U GsgntaA in)t6qu1f4d)Ac n A.M$le Note to SR Q[ Q AchM 3.4.11.1 will be revised to not require the surveillance {Q 3.4u.4.l j %. performance if the block valve (s) is closed DefjC di 'on 4 O. o ce r the ock alv s) r ved n ' Re ir Acti s B.2 nd E .t sur eill ce an t t. ive he wo ing an " met' to " rf rme i t .t wordi of S .4 .1 re sed o c e I.he C ition and exc tion.Dhis change is consistent with traveler WOG 87. mseAT W 9O 14.3.4.18-4] fAnoteisaddedtoITS3.4.8 ACTIONS.indicatingthat] 3.4 48 entry into MODE 5 Loops Not Filled from MODE 5 Loops T Filled is not permitted while LCO 3.4.8 is not met. The addition of this note is based on the performance of a L plant specific LCO 3.0.4 matrix (see CN 1 02 LS 1 of QheCTS3/4.0 package). leses.T ca.";4., @ % -Ik 3.4 49 LCO 3.4.12. "[LTOP] System." provides four differenct methods for pressure relief. Any of the four may be used. However. Eurveillance Requirement 3.4.12.5 requires testing whether or not the equipment is being credited to ineet the LCO. The proposed change adds the word ps42-0 " required" to the Surveillance to exempt its performance gmmp if the equipment to be tested is not being used to meet ~ g the LCO} In addition. SEbecitorial changeWmade. The LCO requirement presentation was clarified.fAlso tobe} the Not o .4.1 w re se or ace'regoir it reg 're ob per orme sinc pie f g,j g fo d" me atu i appr riate erjf. c sist t ith he his an is nsist t /ith ravel r hF-2f6 [Q s.4.n2 1 { n WCGS-Differencesfrom NUREG-1431 - ITS 3.4 9 S/lS/97 l l

- ~. ~. - -. ~. -. -. . - ~, INSERT'6A 9c 0 3.4.12-1 This change is. consistent with th CTS which does not require this surveillance unless the vents'are being used.for overpressure protection. . INSERT 6A-9d 0 3.4.12-1 The CTS surveillance that converts to SR 3.4.12.8 requ' ires that the testing be performed within~ 31 days prior to entering a condition in which the PORV is required to be operable. In other words, the CTS requires that the PORV actuation channel meet the requirement (i.e., be capable of successfully passing the SR) and that the SR be successfully performed prior to entering a conuition in which the PORV is required to be operable. The.STS note for SR 3.4.12.8 allows that the SR does not need to be met (i.e., be capable of being successfully performed) until 12 hours af ter entering a condition in which the. PORV is required to be operable. Since the-SR does not need to be met, neither does it need to be performed until sometime during the 12 hours. - The- . licensees agree that the SR does not need to be performed until-the 12 hours allowed by the STS.and submitted a less restrictive DOC (9-12-LS) and NSHCL (LS-20) to describe and justify the change. The licensees do not agree that it is acceptable to allow.that. the SR not be net upon entering the applicability of the LCO and therefore retain this CTS requirement. In order to retain the CTS requirement that the SR be met upon entering a condition-in which the PORV_is required, the words in the ITS SR 3.8.12.8 note are changed from "Not required to be met. ." to "Not required to be performed. I i i 1: l. L -.~

l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-011 APPLICABILITY: WC REQUEST: JFD 3.4-43 incorporated the CTS 3.2.5 specification for RCS Flow Rate into ITS 3.4.1. Upon further review of ITS 3.4.1, it was determined that incorporation of the CTS into the ITS was not in the correct format per NUMARC 93-03, ' Writer's Guide for the Restructured Technical Specifications." This licensee identified item revises ITS 3.4.1 to correctly incorporate the CTS. ATTACHED PAGES: Attachment No.10, CTS 3/4.4 - ITS 3.4 Encl. 5A 3.4-1, 3.4-2, 3.4-3 Encl. SB B 3.4-3, B 3.4-4, B 3.4-5, B 3.4-6 l l 1 l f

l RCS Pressura, Temperature, and Flow DNB Limits 3.4.1 l 3.4 REACTOR COOLANT SYSTEM (RCS) l 3.4.1 RCS Pressure Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below: a. Pressurizer pressure a EE209} 2220 psig: ggg l b. RCS average temperature s-E581-} 590]'F: and $$$jigsi$ c. RCS total flow rate a [204.000] 371xf104 gpm. pgggps g APPLICABILITY: MODE 1. ...........................-N0TES 11 Pressurizer pressure limit does not apply during :

a. THERMAL POWER ramp > 5% RTP per minute: or 1

b. THERMAL POWER step > 10% RTP.

mW& WWtgegju1i11gfdCon$tfioyA pdfrQlWC.3.4.oti l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l One or w rc ( M A. S DNB A.1 Restore RCS DNB 2 hours $314f43t ar parameter (s) to within 4 limit. s l not within limits. I- - - NOT E - - - l [ t4et appucaWe4> GC5 ^l WC. 3.4 -ot t ( (continued) +cA>Q. % <he.. l WCGS-Mark-up ofhTJREG-1431 - ITS 3.4 3.4 1 S/158 7

. = -. _. l RCS Pressure. Temperature, and Flow DNB L10its 3.4.1 ACTIONS (continued) C0lOITION REQUIRED ACTION COMPLETION TIME C#. Required Action and B.1 Be in H0DE 2. 6 hours g g g" associated Completion TimeUEF18liiillft26hr. not 1

34 *"l 1

met. gg h rVE: -NCI" 2 Restbrb hrainiTate 2%urs ilmestreuw vis; to:wttMin31gts~ MseB l.eM CdhSft1tBiS9f3Mhtit4Fs~ E d.

=="- --del'-WiB g ~ V:2'1 Reduce]}iERNAL_; POWER _to 2 hours RgMow tate _not;within CSOYRIP; Itaits ale 13.12 ReduceW,,,,R, p!gle Khouts N M tfi#i M 'E q j[ph ten:vetp6Tntow g55rRTP; S 5t23 Reduce:3;iERnatP0wER:to 7tnours S MIRTP. 8tB (contiriued) i 1 WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 2 5/158 7 l i

RCS Pressure. Temperature, and Flow DE L101ts 3.4.1 r ACTIONS (continued) COWITION REQUIRED ACTION COMPLETION TIME [tocs.4 on l z -== m. . - ~,; ; _p.. ;,y .. w z... Eshn==c...;=- -+1x MlDM = a.E.# & -'., c 5 Er M S35. M !!Dl5P " " BB 50m M Mitsur#Mrm WEMOllMB N WM .w 4 4 WCGS-Mark-up ofNUREG-1431-ITS3.4 2.4-3 5/15M7

RCS Pressure. Temperature and Flow DNB Lisits B 3.4.1 T BASES LCO M7Any-fou11ng that might bias the flow rate measurement (continued) gat;- t.p. [0.t]t ;-. k.3tejt-d byg..ite.}[.; ;{t.].pa;j;-[a;

r.... r....

.~ r. _...... -. - ..... ~... ~... _,... ~ fewWe shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned tu eliminate the foujing. The LCO numerical values for pressure, temperature, and flow rat; ;.; gi ar. fer t h = ; n 7.t 1;;;tien kt have-not been adjusted for instrument error. APPLICABILITY In MODE 1 the limits on pressurizer pressure, RCS coolant average temperature, and RCS tethMflow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DE limited transient. In all other MODES, the power level is low enough that DE is not a concern. l M b been added to pressurizer pressure is not applicable @uring short tem o indicate the limit n d transients such as a THERMAL POWER ramp 4nereese > 5% RTP r m1nute or a 10% { aee. M conditions represent short tem perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased r in exists to offset the t rar ressure .- y +,- - 7 w n ;a =r. m ! M n a r w _ a. -.1% Another set of limits on DE related parameters is provided in SL 2.1.1, " Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Shald ; vi;ietion of thi; LCO eca, tk ;r;r;ter =;t ;Md t.;ther er r.,t = CL =3 kn k;n eneeeded-l 1 (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 3 5/1/5/97

k l RCS Pressure. Temperatura and Flow DIE L1:;1ts B 3.4.1 4 1 a-BASES (continued) ~ c, a ; w A 6 w ea h t g a N a st>* N s *tc*** % p p.on j ACTIONS L.1 na,;w toes not =ppig 4e eats Wa 2 RCS pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters within LCO limits, action must be taken to restore parameter (s). .00 t;t;l C ri ret; b ret ; cer.tr:l hti; g r; tcr.r.d u ret a g;t;d t; ary i rin; ;ter.dy ;t;t; e g retic.. If t k ir.di;;t;d.""., Otel n c; r;t b kh.; tM LOO li;it, --nzr ;;;t b; red ;;d, ;; r;;;; red by j "; quired fa: tun ::.t. t r;;ter; lL"; ;r;M ;r.d cli;irat tk pter.thi Si vn kt Ma ;f t k ;;;ii r.t ;raly S k ;.t. i The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience. i e C f Q f G,; s q lxa4-mu i If Required Action is not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not i apply. To achieve this status, the plant must be brought to at least s') MODE 2 within 6 hours. In MODE 2, the reduced power condition { 1 eliminates the potential for violation of the accident analysis bounds. i The Completion Time of 6 hours is reasonable to reach the required plant conditions in an orderly manner. t wc. 3 4 olij i assam =wwwemamawamemoursammened l ERhyylMjs.tgjiggB)tje.- 1 ah Wt.elimalgpilgit,'ipowetssiatinggiggs' 1 innstn;andiredsoew.motent1.a.Efor giolatiozn gas [elaccid!!CMYsis M! I k M-Minul_%fsE-&_93 ^-- 8 ZRZ __~is-Wi 2'Mctt lEda89.9MD3Elli!Let10pndL._ W.J M ef M i mesatasameckened_ luramannanseastanw,_ M 2 !Lutarge!!yseynttninA._ - we_wtwerim25N!E (continued) WCGS-Mark-ngs ofhTJREG-1431-Bases 3.4 8 3.4 4 S/1/SM7

i RCS Pressure, Temperature and Flow DIE Limits 8 3.4.1 4 4 5 \\ BASES { j p g,_ - a _ _.._ J {W3.4coiI 4 3 .,;L )~.1+ .'[ - [e f 5 .v r; es* { 1...J J < n.a.: " f' il J- . :.i.@ ! J J.;: g {ter.3.4ou] .... - p a,.... r. v. a...; 1 4 .--a.. .g,u.. .g h e i y- . 7 - y ,a g 4 .,.e_ .l ~. _/_.,*... .[. + .._.a.. .[,. fjA~' j 1 ~ 4 j .2 u .,d ". ;.,. L * ! .. ' 21 ' .A ' 1,. - ~ ' I ' ' k. ; 4 f ' ' ' ' i ' ,_'.k';

1'.

4 '~ h ] y, - -.. t s'.. 1 ,i ?.*: ;. x _ ~ ...T _. - ' S '4.. c(.

, a:

uq

n. :.~

r' 'f an..w. c A. .:,. f. l ' 9.. ... 5 f.. 3rq V .?u W W *i v kuhGQWQ1p WW-1:::? WW F.Mg

g.

t ....o- .m. :,. -MX% g, M Y

O, ss...,,., -. r c &. _&.,..... gj l

h a l K w.. A. , ~..,,

y -.- e -.

..w .m n w wy 4 / ~..- c u a,, y, t s.... g t< u. p..t e >._.,>. e 1 a:.. -.. 1 .w : u_ s v.,...,... e q. g.. 2.;..,,. ) y_. -...,.p., %., .a g . g., ... 3 . _g., _. m -... . :.a? D1$:..l.

.n.L. h*.,' ' 'h O.& 0 i

l 'ba ' _-'c' ':;y'_.... {n e,. t b r 4, s_'. g. I ,.J g 3

J

. n -e a.m. : I.,,.<9.m/.

1; x.

m,. =. j _d R j 2. p ~.EM h,, tr f. A.. y 7 's a 4 W 1. 2.n. = :." .i,

i...-

-s

s. s r

..,JS, y.. - '&.e{,W,y. ( -..f 7 ' .t g.q

A 4.pn

,-..:. + 4. m. ..~;i~)yl_l W ' [e l ] MMilEkIREED!iLTwagtq1M oTaba#%Ai-Es-4 i fBNin:msggg3 prod; i 2-inc. S.4.b04.J 4 l hr.3.4-ott t 4 NMIDIROM1M@.t.1MTsWrLdi_g3LW,. M. 1%il.t.,WMtBWMsisittflfGM tisIprM?MFM" MIRT_P., NMMl2!U3FM_.KachiefFis2MR.EiW'Tdelft11thd, .i 9_.: M.t thicconditiertMngIb.iZreducs5TR.CS?.fl._owf.a.i.t.s.pa. C..., 1 B.2. 1 t } j (continued) a WCGS-Mark-up ofNUREG-1431-Bases 3.4 8 3.4 5 SMSM7 4 4

RCS Pressure. Temperature and Flow DE Lisits B 3.4.1 0, BASES WORS S_toh.ti_niN_di. i we. 3.4.co4.T ggg %D]SKtentTnecessaff,{and ~Cortect_edlpr,ior?toiTdnrestgCttid'ITeher gust _ ion; MMMionTir misfifiedibylaTNgi~thitIst2itiisithati'!EI5_14L IWR AtesM.LhateltoJetteduce_dmiottologrlemi!!a_N. d NitliatistiremtsetgLicitemsens 6NrN $ h y vittr'g gll men RBI 3 SURVEILLANCE SR 3.4.1.1 REQUIRENENTS Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits, the 12 hour Surveillance Frequency for pressurizer pressure is sufficient to ensure the pressure can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. SR 3.4.1.2 Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits, the 12 hour Surveillance Frequency for RCS average temperature is sufficient to ensure the temperature can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. (continued) 3'CG.% Mark-ap ofNUREG-1431 - Bases 3.4 8 3.4 6 S 1/58 7

ADDITIONAL INFORMATION COVER SHEET j ADDITIONAL INFORMATION NO: O 3.7.1-4 APPLICABILITY: DC, CP, WC, CA REQUEST: CTS 3.7.1.1 Action a ITS 3.7.1 Action A.1 and A.2 and Table 3.7.1-1 DOC 01-04-LS3 JFD 3.7-01 This change is beyond the scope of a conversion because of the industry travelers referenced in this DOC (WOG-83, Rev 0 and Rev.1) that have not been approved by the NRC. Comment: Withdraw the change or adopt the STS. FLOG RESPONSE: WOG-83 has been approved by the TSTF and is designated as TSTF-235. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-235 was modified from WOG-83 and these modifications have been incorporated into the ITS. Comanche Peak, Wolf Creek, and Callaway continue to pursue the changes proposed by this traveler. Diablo Canyon will no longer pursue this change due to issuance of LAs 125/123 (dated May 28,1998) which revised CTS 3/4.7.1.1, Table 3.7-1, " Maximum Allowable Power Range Neutron Flux High Setpoint With inoperable Steam Line Safety Valves" and the associated Bases. See licensee-initiated number DC 3.7-008 for more details. FLOG RESPONSE: (supplement) The NRC and the Technical Specification Task Force have agreed upon the acceptable changes to be included in this traveler. Most of these changes were incorporated into the FLOG submittals in the initial RAI response as noted above. The final set of needed modifications have been incorporated as noted in the attached pages. For Callaway, these revisions supplement a supplemental response submitted in the November 25,1998 follow-up letter. ATTACHED PAGES: Attachment No.13, CTS 3/4.7 - ITS 3.7 Encl. 2 7-1 Encl. 3A 2 Encl. 3B 1 Encl.4 19,20 Encl.5A Traveler Status page,3.7-1 Encl. 5B B 3.7-6 l

3/4,7 PLANT SYSTEMS on-13-A' QSAL-1} 3/4.7.1 TURBINE CYCLE i SAFETY VALVES LIMITING CONDITION FOR OPERATION i 3.7.1.1 All main steam line Code safety valves associated with each steam j generator shall be OPERABLE P '9 rr$ng: :: gerfed !- T$'r 3.7 2. f.014A3 . = :c - wm MODES 1,2, and 3. fon enj *t* '" }*"*'*h # *"*- A4- {g a.7.i-4 l j APPLICABILIIY: Meev sh bc. M h MTc. ACTIONb LPa* D y[ N J P* T _* # [01,02;LS-1h 3 y, f Q With four r actor c nt loops and associated steam generators in operation with r more main steam line Code safety valves inoperable, operation in MODES 1,2, and 3 may proceed provided. that I-at-Ls,3 Q 3.7.1-4 \\ within 4 hours either the inoperable valve is restored to OPERABLE status or4he Power Pen;:."cr'ren ' != H!::5 N: ?:r:' :is reduced ?Ol W,$.3 P. per Table 3.71;Aind the Power Ran e Neutron Flux High Trip Setpoin i*n 4 i Gs reduced per Table 3.71 withi ours2th ise, be in at least HOT M #* STANDBY we hin the next 6 hours nd in GOLO HOT SHUTDOWN within4he %_,Ar: ol-17-L5h i fe!!c9n; 3.1 hours. 3G o 2. ;.i-4. l ~ 4 ( Jher$rovjsier%3f.SpetificM.A.arfhot appfftable -Q 3. 7. i - L \\ 1

p. _05-kh o 3.7.1-5'}

TNew) With one or more steam generators with less than two MSSVs l OPERABLE, be in at least MODE 3 within 6 hours and in MODE 4

J0106-Ms L

within 12 hours.

- N#

d f SURVEILLANCE REQUIREMENTS 4.7.1,1 No additional requirements other than those reouired by l Specification 4.0.5J Verify each required MSSV lift setpoint per Table 3.7-2 'in accordance with the Inservice Testing Program. Following testing, ["0107 A r Jift setting shall be within i1%6f 1 l a., Wik one. or rnore, steam generatora dim one. M55V o g-o4 ts l Q 3 7.i q. l j seperable. JmcL Moderater "T'arnperaiwet.Gefficge.nb 1 (MTC-) acro or nega+wa at all powerlevr.ls., re.A n i r e i

THERMAL POWER. 40 f.,87 */o RTf= wi% 4 hours.

x._-_-

l 1

i Separate Condition entry is allowed for each MSSV. 7 01-02-LS-l y 1 " Only required to be performed in MODES 1 and 2. j [Q 3.7.1-6 } ~*' onny rec (uhed ih MDDE 1., c, s.7,i-4 } o i.o g - t3 WOLF CREEK - UNIT 1 3/4 7 1 Mark-up of CTS 3M.7 S/lS/97

1 CHANGE ERBER R$tE DESCRIPTION 01-04 LS 3 The CTS allow continued operation with inoperable MSSVs if 'D4k l the power range neutron flux high trips are reduced. rIplustry. traysh!P @G13, AW9. A efktArfftJtef 1.]p% rovided e ~ i a:E(r-143\\ M revised Aul0Ns to require that: 1) the reactor power be MA b3 g oud reduced to compensate for the loss of pressure relief d#~ capacity to a maxista allowable power detemined in accordance with Westinghouse NSAL 94 001 and NRC Infomation Notice 94-60, 2) the power range r.eutron flux high trip setpoint be reduced for inoperable MSSVs if a positive moderator temperature coefficient (MTC) exists at the allowed percent rated thermal power in MODE 1, and 3) the power range neutron flux high trip setpoints be reduced to account for a control rod withdrawal at partial power with more than one MSSV inooperable.,In addition, the completion time for resetting the high flux trips is revised from four hours to Q hours and the ACTION is revised to specifically require an appropriate power reduction within four hours. This is a relaxation since the CTS require the high neutron flux trip lQ Mi-+ } setpoint to be reduced as required within four hours for inoperable MSSVs regardless of the HTC value._j?endi flipproval o raft Re 1 of 83, the hanges p sed in the tr er hav 1ed to r ain the rent TS r rament reset power r neut flux high t etpoint sed o ntmbe f MSSVs rable to maxi allowa power emined accordance h culation r anal to ac nt for West se 94- [001and Inf on Not 94 60. er, the letion time 72 propo by WOG 83 been ret ' a ified sed on low proba ty of an ent occ ing duri is ti nd the provide ficient me to i r the c is in an rly ma without ucing ransien to rror. R tion of CTS requ t for r tting t eactor tri etpoin is eptable se this r irement i re co vati fv fhan the ACTI specifi either t STS or 83, af revi WCGS-Description of Changes to CTS 3N.7 2 5/158 7

l l INSERT 3A-2 0 3.7.1-4 l l Based on Westinghouse Nuclear Safety Advisory Letter, NSAL 94-001, for plants l licensed to operate at partial power levels with a positive MTC, changes are made to require a reduction in the Power Range Neutron Flux-High reactor trip setpoint in_ addition to a reduction in reactor power when the MTC is positive. This is necessary to limit the primary side heat generation that may occur during a RCS heatup event. With a positive MTC a heatup of the coolant will result in a power increase which requires additional steam relieving capacity. Changes are'made to require a reduction in the Power Range Neutron Flux-High reactor trip setpoint in audition to a reduction in the reactor power when I there is more than one inoperable MSSV on any single steam generator. For a reactivity insertion accident such as an uncontrolled RCCA bank withdrawal from a partial power level the reactor power will increase during the transient until a reactor trip occurs on Overtemperature [ Delta-T] or Power Range Neutron Flux-High. With more than one inoperable MSSV on any steam generator the combined steam flow capacity of the inoperable MSSVs and the I turbine may be insufficient in some cases to prevent overpressurization of the l Main Steam System prior to reaching the reactor trip setpoint. The Action for l reducing the Power Range Neutron Flux-High reactor trip setpoint is modified by a footnote to indicate that reducing the setpoint is only required in MODE l 1. In MODES 2 and 3 the reactor protection system trips in CTS 3.3.1 provides sufficient protection. I l l l l i i I b h 2

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.7 Page 1 of 19 TECHNICAL SPECIFICATION CHANGE APPLICABILITY NUEER DESCRIPTION DIABLO CANYON CONANCHE PEAK WDLF CREEK CALLAWAY 01-01 Reference to Table 3.7 2 is deleted from the LCO and YES YES YES YES A moved to the SR (refer to 01-07-A). 01-02 A note is added to allow separate Condition entry YES YES YES YES LSI for each MSSV which allows the full f%r hours for each inoperable MSSV. 01-03 This CPSES specific revision relaxes the as-found NO: YES NO NO LS2 MSSV lift tolerances from i 1% to i 3t. LA 198/107 issued 10/1/95 to relax setpoint (refer also to 01 LS20). 01-04 Revised ACTIONS for inoperable MSSVs: 1) Mtio Medam@ YES YES YES LS3 specifically requires a power reduction within four ' CTS g hours and 2) requires the reactor power neutron flux k= l hightripsetpointtobereducedwithinghours. M.[q3a. q 1-05 The ACTION of the CTS wnich allowed an exception to YES YES YES YES XA TS 3.0.4 is deleted due to the note associated with [ revised SR 4.7.1.1 which allows a MODE change into MODE 3. one of the MODES of APPLICABILITY of the 4 g,g q LCO. 01-06 The new ACTION adds an explicit requirement to be in YES YES YES YES M MODE 3 in 6 hours and MODE 4 in 12 hours if any SG loop has less than two MSSVs operable. This is one hour less than allowed by LCO 3.0.3. WCGS-Conversion Comparison Table-CTS 3M.7 5/1S/97

i IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS a NSHC LS 3 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE j REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS l The CTS allow continued operation with inoperable MSSVs if the power range neutron flux high reactor trip setpoints are reduced. The amount of reduction in the trip 4 j setpoint is dependent on the total number of inoperable MSSVs per SGand is intended to compensate for the lost relief capacity (heat removal capability and thus overpressure protection) should a transient requiring their operation occur. In the proposed specification, the CTS requirement to reduce the power range hilh neutron flux reactor trip setpoint is retained; however, the time to complete resetting the trip setpoints would be changed from four to satT 4h [- E-3 pa.u-41 The CTS require that, if the MSSV cannot be restored to an OPERABLE status within four hours the power range high neutron flux reactor trip setpoints must be reset in the same 4 hour period. NUREG 1431 requires that the reactor power be reduced in four hours if the MSSV cannot be returned to an OPERABLE status; however, NUREG-1431 would 4 not require resetting the power range neutron flux high setpoints. The Westinghouse j Owners Group (WOG) has proposed changes to NUREG 1431 ftpa9eJer,,Wa34yavrpnaea_ tped$ dP ft1pWU)that: 1) propose that the completion time for resetting the g w. j J power range neutron flux high trS setpoint to compensate for a positive MTC or a 2'sQ control rod withdrawal event at partial reactor power to be@ hours. 2) specifies that power level reductions be per the Westinghouse Nuclear Safety Advisory Letter._ i NSAL 94 01 and 3) deletes the Maximum Allowable

RTP for 5 MSSVs OPERABLE.

ver, pu-9 3 pend 1 appro of dr Rev. 1 WLxi 83 he ch s pro sed in tra er hav ~ a modif to re n the c ent TS ir to re the r ra neut ux hi trip se ints ba on t umber HSSVs perab to a imum i all le power etermi in acco ance wi calcu inns analv s to a unt fors Q inghous SAL 94-0 A. allowed Completion Time to reduce the Power Range Neutron ~rlux tr1p setpoints is reasonable based on operating experience to accomplish the required actions in an orderly manner. The power levels specified per NSAL 94 001 l are based on a conservative algorithm developed by Westinghouse to bound the required relief capacity, sTn-23D %s.04} The above changes are consistent with NUREG 1431 as revised by and NSAL 94 001. 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 sat forth in 10 CFR 50.92(c) as quoted below: "The Comission may make a final determination. pursuant to the procedures in 50.91. that a proposed amendnent 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 facility in accordance with the proposed amendnent would not: WCGS-NSHCs - CTS 3M.7 19 S/158 7

l INSERT 4-19 0 3.7.1-4 1 Based on Westinghouse Nuclear Safety Advisory Letter, NSAL 94-001, for plants licensed to operate at partial power levels with a positive MTC. changes are made to require a reduction in the Power Range Neutron Flux-High reactor trip setpoint in addition to a reduction in reactor power when the MTC is positive. This is necessary to limit the primary side heat generation that may occur during a RCS heatup event. With a positive MTC a heatup of the coolant will result in a power increase which requires additional steam relieving capacity. Changes are made to require a reduction in the Power Range Neutron Flux-High reactor trip setpoint in addition to a reduction in the reactor power when there is more than one inoperable MSSV on any single steam generator. For a reactivity insertion accident such as an uncontrolled RCCA bank withdrawal from a partial power level the reactor power will increase during the transient until a reactor trip occurs on Overtemperature [ Delta-T] or Power Range Neutron Flux-High. With more than one inoperable MSSV on any steam generator the combined steam flow capacity of the inoperable MSSVs and the turbine may be insufficient in some cases to prevent overpressurization of the Main Steam System prior to reaching the reactor trip setpoint. The Action for reducing the Power Range Neutron Flux-High reactor trip setpoint is modified by a footnote to indicate that reducing the setpoint is only required in MODE 1. In MODES 2 and 3 the reactor protection system trips in CTS 3.3.1 provides sufficient protection. i

1 IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 3 (continued) 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2. ' Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3. Involve a significant reduction in a margin of safety." The following evaluation is provided for the three categodes of the significant hazards consideration standards: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed change does not result in.any hardware changes or changes to operating ) methodologies. This revision does not affect an accident initiator of any analyzed accident. The proposed change is based on recognition that 1) Reactor power must be limited to compensate for lost pressure relief (heat removal capability) with inoperable MSSVs, and 2) that there is a need to prevent a reactor power increase via { a reduced power range neutron flux high trip setting, for transients that would cause added heat transfer to the secondary system. The proposed change also allows for an { appropriate amount of time to 1) restore the MSSV to OPERABLE or reduce power and 2) reduce the power range neutron flux high reactor trip setpoint The probability of an accident occurring and requiring MSSV operation during the hour Completion Time would be very small. The time extension allows the reactor r p setpoint reduction to be performed in a more deliberate manner, thereby reducing he potential for inadvertent reactor trips introduced during the setpoint r uction. However, because the accident analyses are unaffected by the proposed change, the consequences of the accident analyses are not affected by this change. a p3a.i_4 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes do not require physical alteration to any plant system or change the method by which any safety related system performs its function. No changes in plant operation result from the changes, and no new event initiators are introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. WCGS-NSHCs-CTS 3M.7 20 $/1587

l Industry Travelers Applicable to Section 3.7 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-36, Rev. 2 Incorporated 3.7-42; only Adds NOTE that states that LCO applicable to 3.0.3 is not applicable. DCPP TSTF-51 Not Not Applicable Requires plant specific re-analysis Incorporated to establish decay time dependence for feel handling accident. TSTF-70, Rev.1 Not Not Applicable LhC approveddpa.u 4 Incorporated WC cJ.offM TSTF-100 Incorporated 3.7-05 and 3.7-19 NRC approved. TSTF-101 Incorporated 3.7-29 NRC approved. TSTF-13@) Incorporated Not Applicable @RC agreQ { TA 3.1-a:f} TSTF-140, Rev.1 Not Not Applicable Not NRC approved as of Incorporated traveler cut-off date. / sti b W Incorporated 3.7-01 Q 33.,-4 Cmy-2m 2p g 3,7 37 p,,3,,,...,3,,,,3.2 t) H a Incorporated 3.7-56 14 57.2-3\\ @3_T,3 inc-puht % Apehot4. mc vr>MD ra. s.,-oorj

p.,%~,w m ynca g m 2,-oomt ea.

MSSVs 3.7.1 3.7 PLANT SYSTEMS 1 3.7.1' Main Steam Safety Valves f.MSSVs) LC0 3.7.1 The-FJij MSSVs ggghigenapBprishall be OPERABLE-ee Op;;ifi;d ir. T;ti; 3.7.1 1 cr.d Tebi; 3.7.12. $M. N@ 8. l APPLICABILITY: MODES 1. 2. and 3. ACTIONS ..................................... NOTE Separate Condition entry is allowed for each MSSV. i CONDITION REQUIRED ACTION COMPLETION TIME (B $3a.i-4) S g (One or mor; i g ired p.1 Reduc; p;;;r E R W. 4 hours SSVs,,Arvum3idE) I M to less than or wsm gssgers $4% equal to th; ;,,,,lic;bi; cia d.... h or mere Mssvs Ib.x_ i m :{ t RTP ihoperab\\e.. N in

3 Table 3.7.1 1W'11he m - _ - - -

ww1mi2mgE

*w----

N e=r*-W" MSSVs. 3 <*1ade4 a 3.n c,wc3 wim one. cal b M55V inoperabic. Omd ,[D # E g "- we. nic. p.sh at ("*3 P* " " D h p' m 3 gg Neut0MEBilHIRB i r e gijn g a n ge g nt E t Jo f!BL M E W 81 \\q3,u_;j toTthKEEmma M agen s mage MR@!BBtE!ER!F1 3273sterwnstber of] M ^ f$, omor m~ ore steam A (continued) A.I hauu. w Riw.powsR gener>tx s wah one. b 6..Es 13 -/. arr. W5 ) i nssy inoperable. M*c. Moderabv-Temp 4atwu. _J CoeEEteny MTc4 we j 421.i-q j devch.ptWt. at alt fotgr ce neo f WCGS-Markap ofNUREG-1431 -IIS3.7 3.7 1 5/1587 ~

MSSVs B 3.7.1 BASES ACTIONS A.1'antii f6oDtufu'ed) l93.7.1-4\\ ~

p. N nn.,rn,

. r,aw __;..__a ,,,rn o., 2_ .u. m _ _ _.. _ _. _____ e . b 2-L g. ...bsvtrw. I wn b. . g. v b. W usi b.gb b y.s.ub5Ib Vpi b. V. 5 b yVvyw. j 1;;;l si;;t r ther, the reli;'_;;pecity s' the r; ;;;ir.ir; "S"'h. 1 _J... J "PY. fen.u A I n. nt_, fan J_ J_m__ .'__J

2. w_1 1. w_.

J .wgswwbu . shst nh vvvb.g id wb bb Pli u sww rv d e i 4 o where:- i n,n

n.._.2.. _ _ ;,,,rn.o,

nn,_,rn, .u ___,,,_2.,__.... __ m. ...m m .-m .... ~.. ~.,,..

,;.reter grc'; c' ; ; p;r;;r.t c' ",";

l %, u u, _ _ _ _ _ _ _ _ i. t. e m, u _ u..r ru. e,,- .y, . r. .g.....~ ....,~. ~., m, i r; p;;ti;;1y, 4 n,,. u uee,, 2.._. nnen.n. r w e,

b. N. ~.

Wu 6.v u#6. 1 . u. u.d' F1., 2.._. 2.._~ys. _._ L S. 22 .. ~..~ i 3 t"D../.,*

t. 'D.,P -

.L_ - _, J. _ r. ___-_J. _t .L-9 6 e ..ro m..,.. u..e ru. .~ v. .. ~. ~.. m. _.-..._1.,. .J _ 2 2. _~_ o.s._ ______.2 _L_. us cs.i y s. 2. - a 2., e {Q 5a.1 se ) c g.1 and f.2 i J_ 4he. bqmred Achim xe.dcomeb@- I fem.u _ ueru._ ___m.._._._.m mnen. v. n, e _.... .c 1-L_ .~. ~..... ~, -. ~m o I .s nofiT_~[ Q P g RangeWR!utron F N - Hign+ W setpoi 24 within the associated completion Time, er if one or more steam 'gDA Mssy, 9enerators have.(] pesthefWoJ85M)er>Je9H0, the unit must be placed 3 o in a MODE in wh'ich the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in on orderly manner and without challenging unit systems, i 4 i i 1 (continued) WCGS-Mark-up ofNUREG-H31 - Bases 3. 7 B 3.7 6 5/15/97 i,

l INSERT B 3.7-6 0 3.7.1-4 J In the case of multiple inoperable MSSVs on one or more steam generators, with a reactor power reduction alone there may be insufficient total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude l overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. Furthermore, for a single inoperable MSSV on one or more steam generators when the Moderator Temperature Coefficient is positive the reactor power may increase as a result of an RCS heatup event such that flow l capacity of the remaining OPERABLE MSSVs is insufficient. f!".crefore, 4-reddition to Required ^.ction B.1, -"ich cpeci#ies ar appropriate reduction in reacter p0:er withia hourc, P.equired ^.cti0r B.? cpeci#iec that the Power A (Range Neutrer Ou,,_uigh reactor trip cetpcint be educed within 72 heurc. 43 IThe 4 hour Completion Time for Required Action B.1 is consistent with A.1. An additional 32 hours is allowed in Required Action B. 2 to reduce the setpoints. The Completion Time of 36 hours is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, operating experience in resetting all channels of protective ' function and on the low probability of the occurrence of a transient that gld result in steam generator overpressure during this period. l A sensitivity study (Ref. 7) was performed to analyze the loss of load / turbine trip event ineiated f rom power levels based on Table 3.7.1-1 and assuming both beginning of life and end of life reactivity feedback conditions. The results of all cases studied showed that the secondary system peak pressure l was maintained below 110% of the secondary system design pressure limit. l Required Action B.2 is modified by a Note, indicating that the Power Range l Neutron Flux-High reactor trip setpoint reduction is only required in MODE 1. l In MODES 2 an 3 the reactor protection system trips specified in LCO 3.3.1, l " Reactor Trip System Instrumentation." provides sufficient protection. The allowed Completion Times are reasonable based on operating experience to l accomplish the Required Actions in an orderly manner without challenging unit l systems. l l l l l

l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.7.10-14 APPLICABILITY: WC, CA, CP REQUEST: CTS 3.7.6 Action b, Modes 1,2,3 & 4; Action c, Modes 5 8 6 ITS 3.7.10 Action B, C, D, E, and F l DOC 10-20-L S39 JFD 3.7-57 These changes are beyond the scope of a conversion because the industry traveler referenced in this DOC (WOG-86) has not been approved by the NRC. In addition, for CPSES, the Action D condition statement appears to be contradictory because with two inoperable trains, it is not clear how any pressurization occurs. l l Comment: Withdraw the changes or adopt the STS. FLOG RESPONSE: WOG-86 has been approved by the TSTF and is designated as TSTF-287. This traveler has been submitted to the NRC and is under review. l The proposed wording in TSTF-287 was modified from WOG-86, and l these modifications have been incorporated into the ITS. This results in changes only to the Bases for the Required Actions for ITS 3.7.10. The FLOG continues to pursue the changes proposed by this traveler. This TSTF is not applicable to DCPP but is applicable to CPSES. t For CPSES, Condition D refers to two CREFS trains inoperable due to inoperable CR boundary. The CR boundary inoperablity is usually the result of boundary degradation such that with either CREFS trains running the system is unable to maintain a pressurization of >.125 inches water l gauge at 800 cfm. Action D merely requires repair of the degraded I. pressure boundary. j FLOG RESPONSE: (supplement) The NRC and the Technical Specification Task Force have agreed upon the acceptable changes to be included in this traveler. Most of these changes were incorporated into the FLOG submittals in the initial I RAI response as noted above. The final set of needed modifications have been incorporated as noted in the attached pages. ATTACHED PAGES: Attachment No.13, CTS 3/4.7 - ITS 3.7 Encl.5A Traveler Status page,3.7-23, 3.7-25 Encl. 5B B 3.7-65, B 3.7-66, B 3.7-67 i 1

Industry Travelers Applicable to Section 3.7 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-36, Rev. 2 Incorporated 3.7-42; only Adds NOTE that states that LCO applicable to 3.0.3 is not applicable. DCPP TSTF-51 Not Not Applicable Requires plant specific re-analysis Incorporated to establish decay time dependence for fuel handling accident. TSTF-70, Rev.1 Not Not Applicable @NRC approveddl re u.accl l Incorporated

g e s.ogg g TSTF-100 Incorporated 3.7-05 and 3.7-19 NRC approved.

TSTF-101 Incorporated 3.7-29 NRC approved. TSTF-13@)Incorponted Not Applicable @RCAjyroQ l rA.5.7-a:sj TSTF-140, Rev.1 Not Not Applicable Not NRC approved as of Incorporated traveler cut-oK date. \\ / nti b i W Incorporated 3.7-01 i Q 2.~t I-4 w TD -~ ' Incorporated 3.7-57 ( [ [@3~i80+.31831" Ll MM Incorporated 3.7-56 I4 5M.2-3\\ i (npr F-113 IMCo'T*'ded Q ApplacaW WC ypetautd TR. 3.7-col l 974r Gym e wn e y n n-oezi O s

~_.__ L l-i CREVS%REFS 3.7.10 i L

3.7 PLANT SYSTEMS I

3.7.10 Control Room Emergency Tiltr;tica Ventilation Syst= (C"ir3 (CREVS) M$N LCO 3.7.10 Two-C-REFS CREVS trains shall be OPERABLE. g w w p 7he. conMot naam ipe.Andary my kNr epansA.jnterMate Qu. nae, o.a min.. Path comb is,. \\Q3I,D,i+} APPLICABILITY: MODES 1, 2, 3, 4,T D E, $Affd During movement of irradiated fuel assembliesr. ~.. . iiia; 00"i AL""ATIO"O ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME i l .A. One CaiTS CRQ(S, train A.1 Restore Cair CREVS 7 days -inoperable. train to OPERABLE ' status. 'i B. T h 3BEMS_tralas B.1 Restore control: room 24, hours MJER'ID IgtBfy;f.o_(LPERABLE EINy,d.$ 13lLRElBTc5fftfolroom status. _W p (in moor.s i,2.s_,anA47 i

80. Required Action and BC.1 Be in H0DE 3.

6 hours associated Completion M5ka Time of Condition A or B ANIl L not met in H00E 1, 2, 3, or 4. BC.2 Be in MODE 5. 36 hours l l (continued) l WCGS-Markup ofNUREG-1431 -I153.7 3.7 23 S/1587

N 3.7.10 4 ACTIONS (continued) CO M TION REQUIRED ACTION COMPLETION TIME yp s.r.io -14 1 @r - ~' M M3 M3s M. 3bEil@DiER g!ERATICE yJap*pyy s c.. *, - 851 F%*^fJk BE.2 Suspend movement of Immediately irradiated fuel g assemblies. &:- Two -GREFS-GES trains inoperable M N during movement 2n-sr of irradiated fuel ld5 $- 54-} AbTERATf0NS. s EF. Two -GREFS-GEUS trains EE.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2 k!@@.dN5$ 3, or - wLROJ 9M M. SURVEILLANCE REQUIREMEhTS SURVEILLANCE FREQUENCY i@!$3M SR 3.7.10.1 Operate each C"H S CREVS train pre gutit.8 lt on 31 days fgttiGTrift for $1tRicpigi;riuous'hourt'iFrfttittfis ggggggg W.WEr er 'fer ;ystera witt.e 4 ~ ~ " tutors) uriitasr @gLrm.rainliltratloKJJ1ter g"ggg g ga g (continued) WCGS-Markup ofNUREG-101 -ITS3.7 3.7 25 S/158 7

GREfS-CREVS 1 B 3.7.10 i i l l BASES i LCO The GREfS-CREVS is considered OPERABLE when the individual components (continued) necessary to limit operator exposure are OPERABLE in both trains. A-GREFS-CREVS train is OPERABLE when the associated: a. ';r, t ",ecirculation and pressurization fans 1are OPERABLE: i-l b. HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their L filtration functions: and c. Heater,Leoisture separator.; d;.htcr. ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained. l l d~' Contro17Roon~ Air Conditioning; flow; path int.egrity'.is maint_ained. In adoition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors (russar a s.1-6DlGC 10-l4- \\ Note;that the ControlfRoom~A_ir, Conditioning Sys_tes]CRACSEKrasj i @ system _,tpltheiC_REVSRTptelCREVSirejsains; c8p_able;o1"perftirshis"!1ts saTetyJunctionJtofidedithECMiaitMow? path))jLW~.MfDMt circulation;canibe; maintained 73 solation or1WMtWERRCSMt finPath~ canla]soirenderith. CREVSifldhCpath11nograbje;.7;'t_hes~ e e situations.1.00s; 3;7J10,Eand ~K7?l1[may, belapplicable, @e du gce== x_-c r W G) }o3.74-1[ APPLICABILITY assemblies [GREM-CREVS must be OPERABLE to control In MODES 1 2, 3, 4,;5Eand 6, and during movement of irradiated fuel l r i during and following a DBA. In H0DE;5~orL6, the GREFS-CREVS is required to cope with the design basjs release from the rupture of er, cd;ide a waste gas tank. During movement of irradiated fuel assemblies ;r.d C0": ',L""',TIZ, the GREFS-CREVS must be OPERADLE to cope with the release from a design ~ basis fuel handling accident. t (continued) d WCGS-Mark-ap ofNUREG-1431 - Bases 3.7 8 3.7 65 5/158 7

I i i INSERT B 3.7-65 0 3.7.10-14 The LCO is modified by a Note allowing the control room boundary to be opened intermittently under administrative controls. For entry and exit through j doors, the administrative control of the opening is performed by the person (s) entering or exiting the area. For other openings these controls consist of l stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated. 1 l l l j l l I-

_ - -. ~. - GREM-CREVS B 3.7.10 i BASES l ACTIONS L1 When one GREE-CREVS train is inoperable, action must be taken to i restore OPERABLE status within 7 days. In this Condition, the remaining OPERABLE GREFSKREVS train is adequate to perform the control room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE GREM-CERS train j could result in loss of GREM-CREVS function. The 7 day Completion l Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability. h=MocE51,2 A A*hlQ33.10-14 } o-CFITAuaE f } qs.7,io-tQ Ifithe controllroom boundary,is'Jnoper,ablisucjh that&CRQS tgejn. i i cannot~establisitor aiaintain' the_reguireid'pressureZactionsisWh l takenitgI W inn W ~ contro]Lrio3m boundaryMthig.24Dsl.: W"2D[our unplition;11me;isMasgnabKnascilwthe'l# pr' bLgGJtylef31M_'pecurring[derifigT.tpis'tigeiperiN2MW o ava1]ab111tyloO:helCREVLto; provide affilter.ed;~ environment]albeit wfKpotentlaEcontroliroomi.inleakage). BC.1 and BC.2 In MODE 1, 2, 3, or 4, if the inoperable GREM-CREVS train orlthe control] room 1 boundary cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on i operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. (continued) l { WCGS-Mark-up ofNUREG-1431 - Bases 3.7 8 3.7 66 5/] sap 7 l l

GREFG CREYS B 3.7.10 BASES ^ D. l. l, D. I.2,) ACTIONS O G0 1.1. and 60.2.2 ~ (continued) In_H0DE,5,or 6, or during movement of irradiated fuel assemblies-ee during CORE ALTERATIONS. if the inoperable GREM-CREVS train cannot be restored to OPERABLE status within the required Completion Time, action must be taken to imediately place the OPERABLE GREF&-Cl(VS train in the scrgcncy CRVIS mode. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur. and that any active failure would be readily detected.f r An alternative to Required Action 60,1 is to imediately suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position. Rcquir;d Action C.1 i; ;;dified by ; Not; indicating to plecc the

y;tc; in th; toxic ge; pretection ;;d; if Oute;; tic tron;fer to toxic ;;; protcction med; i; ineperabic.

BE.1 and DE.2 In H0DE 5 or 6, or during movement of irradiated fuel assemblies--ee during 00"E ALTE"ATIONS. with two GREM-CREVS trains inoperableffT "rb ns-ne ha ue 1 ra tyd1 u _f J .e., j a. _.... oni J1.t a a'ry no re or ,.1 O (C'~ le n~ ~l ad y~ i on action must be taken 10 m0-841 a immediately to suspend activities that could result in a release of radioactivity that might enter the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position. JoM.ld-n j fActdn D.l.1 <equires We CREVs irsih placut ih ope <4 tion be capato_ y 4 of be i g po we<=a. by w e.mege,-,3 g.m som m.w.o o m ->, assucs oP5ss24mi i ry of the. c c:aEvs traih A h u.~,n % e met ' of a Cuct H>ndimj Acc ict er t Dw3% k. gh whh p+de co.rmr-re,<t J+b a Los s of c.ff site. poum. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3. 7 B 3.7-67 S/1S/97

l l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.7.13.2-2 APPLICABILITY: WC, CA REQUEST: CTS 3.7.7 Action, insert M[New c]; CTS 3.9.13[new c) ITS 3.7.13 Action B, C, D, E, and [F) DOC (3.7) 10-20-LS39 l DOC (3.9) 12-12-LS26 JFD 3.7-57 These changes are beyond the scope of a conversion because the industry traveler referenced in this DOC (WOG-86) has not been approved by the NRC. Comment: Withdraw the changes or adopt the STS. J 1 FLOG RESPONSE: WOG-86 has been approved by the TSTF and is designated as TSTF-287. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-287 was modified from WOG-86, and these modifications have been incorporated into the ITS. This resulted in l changes only to the Bases for the Required Actions for the Emergency Exhaust System, ITS 3.7.13. The FLOG continues to pursue the changes proposed by this traveler. In addition, due to the change in traveler status, WOG-86 has been replaced with TSTF-287 in DOC (3.7) 10-20-LS39, DOC (3.9) 12-12-LS-26 (for WCGS only), NSHC (3.7) LS-39, NSHC (3.9) LS-26 (for WCGS only), and JFD 3.7-57. Also see the response to Comment Number Q 3.7.10-14. FLOG RESPONSE: (supplement) The NRC and the Technical Specification Task Force have agreed upon the acceptable changes to be included in this traveler. Most of these changes were incorporated into the FLOG submittals in the initial RAI response as noted above. The final set of needed modifications have been incorporated as noted in the attached pages. ATTACHED PAGES: Attachment No.13, CTS 3/4.7 - ITS 3.7 Encl. 5A Traveler Status page,3.7-30 Encl. 5B B 3.7-77 l l l

_ ~ __. _ _ _ __.._ _ _ ____ _ _ _ ___ 4 Industry Travelers Applicable to Section 3.7 4 i j TRAVELER # STATUS DIFFERENCE # COMMENTS i TSTF-36, Rev. 2 Incorporated 3.7-42; only Adds NOTE that states that LCO applicable to 3.0.3 is not applicable. DCPP 4 j TSTF-51 Not Not Applicable Requires plant specific re-analysis Incorporated to establish decay time dependence for fuel handling accident. TSTF-70, Rev.1 Not Not Applicable @NRC approveddj ra.u.ces l Incorporated ~tWeecht effAet% TSTF-100 Incorporated 3.7-05 and 3.7-19 NRC approved. 4 TSTF-101 Incorporated 3.7-29 NRC approved. l TSTF-13h) Incorporated Not Applicable %d ajyrog l rA 3 3-erj f TSTF-140, Rev.1 Not Not Applicable Not NRC approved as of l Incorporated traveler cut-off date. / ntt b k' Incorporated 3.7-01 )_ LkS6D ) 'M ' Incorporated 3.7-57 [oTc e)[@ 3. l to-'+ 3.1.t3 2-Ll ] @-2MM Incorporated 3.7-56 14 5.7.2-3\\ Cn,rg-i,a inco<F dea e4 + Aeekc=A

c ofeMD Ta. s.7-oosl

[sr2- $4 in c.e'P r*td-kb+ APP C M - [ w a1-col ( li

i l EmerDency~ERhaust'SystedFBACS 3.7.13 3.7 PLANT SYSTEMS 3.7.13 ruci Ouilding Air 01:;nup Systa 'ISACS) Emergency Exhaust System (EES) g ps; y l o s~ 1' e.1-2. } l LCO 3.7.13 Two FBAGS EES trains shall be OPERABLE. r e,.pg_ _ __ e We. and hW,,3 6u:itdm er far.\\ bu'iid, boomd$ mu3 be. (.37.s7 (ped giite, rdittent-Ty unde.v t2dmini__ _._ _. - - _ _ _ "c45. tW6 c.ontT ~,.., p g APPLICABILITY: MEES 3Z223Dgf During movement of irradiated fuel assemblies in the fuel i building.

L?MEL.T5,2-?EL -hd, b5i_OTC :i,-if:si2L--ihElJ,5,~M~

~ N a ThtSIS 3adt.nCoperpt.1on.is,J;gatired on11.1u,3XIlEli2ZEBl jgg~g; E319GlEIEgiroY"oper.at1K_ffrinRI112d ntDY31gr1DsMMll ~ SEEEDB3EE3103Ra. !b1LeCBu2HCIIEllBBSimC N 3:29~FI:3ETErsemMEEIEEf?E5MMMFi ACTIONS CONDITION RE0VIRED ACTION COMPLETION TIME A. OneElgFBAGS-train A.1 Restore EES FBAGS train 7 days inoperable in; MODE 1, 2, to OPERABLE status. 7747"m. 3E33 l B.' TWoptta1M B.1 Ristor.OJ1XJ1.lary 24' hours ilSDE!EEJIDEto bhiMiDglou0diry,to it!$fRQREl.11ary OPERABLE status'. sgyy.gl l buildimJamodary in M2E3lt".1. I (continued) Il' CGS-Markup ofNUREG-1431 - ITS 3. 7 3.7 30 S/1S/97

Emergency Exhaust SystemFBAGS l B 3.7.13 BASES l APPLICABLE for a fuel handling accident and for a LOCA. These assumptions and SAFETY ANALYSIES the analysis follow the guidance provided in Regulatory Guides 1.4 (continued) (Ref. 5) and 1.25 (Ref. 4). The ICACS Emergency Exhaust System satisfies Criterion 3 of-the-NRG NRC Policy Stat nint 10~CFR 50.36 (c)(2)(ii). LCO Two independent and redundant trains of the "?ACS Emergency Exhaust System are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure could result in the atmospheric release from the auxiliary. building or fuel hadling building exceeding the the guideline limits of 23 CFR 100 (Ref. 5) limits in the event of a LOCA or fuel handling accident. The ISACS Emergency Exhaust System is considered OPERABLE when the individual components necessary to control releases from_the auxiliary or s p sur; in the fuel handling building are OPERABLE in both trains. An-FBACe Emergency Exhaust System train is considered OPERABLE when its associated: a. Fan is OPERABLE. b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function: and c. Heater, dc;iiistcr. ductwork, volvcs, and dampers ar ' OPERABLE, and air circulation can be maintained. @scxT a 37-D_. h 3.7.13.2-2.j APPLICABILITY In H0DE 1, 2, 3, or 4. the TSACS Emergency Exhaust System is required to be OPERABLE in the SIS mode of operation to provide fission product removal associated with potential-EGGS radioactivity leaks during the post LOCA recirculation phase of ECCS operatior, duc to a LOCA ad 1;;kagc fr a anteir. cat ad nr.ulus. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3. 7 B 3.7-77 S/15/97 l

_._v. _.- ~. __ _ _ _ _ _ ..._.._____m.. ?- INSERT B 3.7 0 3.7.13-2 .The LC0_is modified by 'ar Note allowing the auxliary or fuel building boundary g - to be opened intermittently under administrative controls. For. entry'and exit through doors, the administrative control of the-opening is performed by the. [, _ person (s) entering or exiting the area. For other openings these controls. consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a. l! method to-rapidly close.the opening when a need for auxiliary building or fuel-building isolation is~ indicated. s 1 e i i b I l. i' i t t { y N -p,d y y, p y-y am;< ,..my--y ---c'*?---q+ --f t' r

l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.9-7 APPLICABILITY: DC, CP, WC, CA 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) l 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: (original) 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 j retain this change within the submittal pending NRC review of the proposed traveler. Additionalinformation supporting this DOC is in NSHC LS-20 in Enclosure

4. DOC 4-10-LS-20 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 j 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 l containment provided that specified administrative controls were l 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 isciate 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 l l

_ _ _.. _. _ _ -. _ = - - l, the requir::mants for airlocks that allow both doors of an airlock to be open during CORE ALTERATIONS and during movement of irradiated fuel inside containment provided that administrative controls are in place to quickly close one door and establish containment closure. i. 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 - 6 containment closure function and will serve to limit the consequences of a fuel handling accident as described in the Safety Analysis Report such that the results of the analyses in the Safety Anaiysis Report remain bounding. In considering the consequences of a design basis fuel handling accidents inside containment, the assumptions in the analysis l 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." ? I For DCPP, a preliminary 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 L consequences. f FLOG RESPONSE: (supplement) The proposed changes to CTS 3.9.4 and ITS 3.9.4 were based on traveler WOG-76. WOG-76 has recently been designated L TSTF-312 and transmitted to the NRC in December 1998. ' Based on the present status of the generic change process for the STS, it appears that travelers TSTF-312 will not be approved by the NRC in time to support the initial license amendments for the FLOG plants. In order to facilitate the issuance of these initial license amendments, an alternate approach has been developed which relies on the CTS, plant-specific information, and/or the NUREG but does not rely upon the traveler. This alternate approach is hereby provided as an interim submittal to allow issuance of the initial license amendments. The changes which rely upon the traveler L can be processed in subsequent license amendments following approval of the traveler by the NRC. The CTS and ITS have been modified to remove TSTF-312, which will result in DOC 4-10-LS-20 and JFD 3.9-11 as being "Not Used." For Wolf Creek, a Note to LCO 3.9.4 will be maintained to adopt the CTS wording approved by Amendment No.107, dated July 11,1997, that allowed l penetrations P-63 and P-98 to be opened under administrative controls. i Additionally, SR 3.9.4.1 has been revised to except containment penetration P-63 and P-98 if they are open under administrative controls to comply with the ITS format and rules or usage (e.g., SR 3.6.3.3 and associated Bases). New JFD 3.9-16 has been initiated to address the Wolf Creek specific changes to adopt CTS. !~

-. ~ l l-ATTACHED PAGES: l ' Attachment No 13 CTS 3/4 9. - ITS 3.9 l Encl. 2 9-4 Encl. 3A 5 l Encl. 3B 3 L Encl. 4 - 1,44,45,46 I Encl.5A Traveler Status page, 3.9-5, 3.9-6 Encl. 58 B 3.9-13, B 3.9-14, B 3.9-15, B 3.9-16, B 3.9-17 Encl. 6A 2, 3 Encl. 68 -3 t l i i f f

- ~ ~ ^ ~ ^ ^ ~ ~ ~ ~~~~~~D J { Q 3.9 ~I \\ ~ ' I.) f s rom.sW A f 1 A, b h S... W. ~ r'r g n u_^ "6_. 'O arirawn T - M * *" " " "3 .-2 REFUF1 ING OPERATIONS A, -, ',..i, rG i -vu i e nb i - N) % "*- L i _ _~ ~ w m m. n i sw arn vs. c. Won s.' 3/4 9.4 CONTAINMFNT BUILDING PENETRATIONS N'Y ~ _ _ n_ LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status: i j

a. The equipment door closed and held in place by a minimum of four

bolts,

b. A meimum of one door in the emergency airlock is closed *, and one door i

i in the personnel airlock is capable of being closed, and IG 3 9 ~~l I l

c. Each penetration providing direct access from the containmer atmosphere to the outside atmosphere shall be either:

l**"% 4 1 i

1) Closed by an isolabon valve, blind flange, or manual valve, or i

approved funcbonal equivalent, or l

2) Se capable of being closed by an OPERABLE automabs containment RF4#44sI@

purge isolabon valve. -dra _- A W.3.5-004) 2 ICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. l d 3.1 J ] ACT.10N: l With the requirements of the above specification not satisfied, immediately suspend all opershons involving CORE ALTERATIONS or movement of irradiated fuelin the containment building. 3 ) SURVEILLANCE REQUIREMENTS 4 4.9.4 E,,,ch of the above required containment building penetration shallbe - M3$-7 determined to be :5.:: L-5 :: ; fed ::n "*J:c er ;:i': ef i::n; ted by M-n -- C"EP_^" " ..:^ : ---'in.;n: ;"r;: ':: 'd :.J.".:- 1T h~. Mer $qq,ghg,g,4q '- S :^_. ':nf -'. '::' :nx ; r fr;; dunn; CO".5 ^L"_P^."ONS ^ en-;n' 3etenenM eN-' % M nn'_:. c:n' br"f! ; bygn the required status at least) (once per 7 days)

. '!:-_i.: 5 : x _ _

x cre !- t. fr;9:d xnf!"_:n, Worify each ] l^*4034151 .(--b. :Edsting the containment purge is@,a valves per the applicablerequired containment p " " ^ ~ " " ^ " ^ ' ^ " portions of Specification 4.6.3.2(at least once per 18 months, by use of an) { actual or simulated signal) { [((

An emergency escape hatch temporary closure device is an acceptable replacement for the airlock door, 5 --

g33 -l \\ { ce th( pro ing ta ~ i ont os ide os re y be ni ted i S u i Is. ) N ___z WOLF CREEK - UNIT 1 3/4 9-4 Amendment No. 74,89 (Next page is 3/4 9-9) Mark-up ofCTS3M.9 5/158 7

i CHANGE NIABER lEC DESCRIPTION 4 03 LS 5 l' The frequency of verifying that [ containment ventilation 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. he new frequency is consistent with those SRs applicable to ESFAS type functions and inservice valve testing which are appropriate for the containment isolation function. % N5EstT 5A -- DID-i Q 3.9-eq 4-04 TR 1 Revised Surveillance requirement to allow for increased flexibility in using an actual or simulated actuation i signal. Identification of specific signal is moved to the Bases. 4 05 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). l 4 06 LS 23 Not applicable to WCGS. See Conversion Comparison Table i (Enclosure 38). l l 4 07 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). l 4 08 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). l 4 09 LS 14 Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 38). i h 4 10 Adds a footno stating th penet ion f1 athsthat] l provide t access the ntainee atmosphe to l g g, the o ide atmos e may unisol under i a nistrative trols. is c is co stent h 4 l raveler 6 and prev sly app ed l adminis ative co ol fo air locks. portonnel,M 93 A-7[ l ~ SEtTse!CA - 5 01 R Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 6 01 R Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 7 01 R Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). WCGS-Description of Changes to CTS 3N.9 S S/15/97

hRT3A-5 w 0 3.9-7 Q3 4 *l A note is added to LCO 3.9.4.c and the 7-day SR to state that l containment penetrations that provide direct access from the contai ent l {atmosphereto.theoutsideatmospheremaybeopenunderadministr ve l controls. The note would allow these penetrations to be uniso ted l during CORE ALTERATIONS and movement of irradiated fuel ass lies l within containment provided that specified administrative ontrols were l empl oyed. The proposed Note is acceptable based on ad 1strative } controls that consist of written procedures that reg re designated personnel having knowledge of the open status of t valves in question and specified persons designated and readily av able to isolate the open penetration in the event of a fuel handl g accident. These administrative controls provide protection uivalent to that afforded by the administrative controls used to e ablish containment closure for a containment personnel airlock. The C staff has allowed changes to the requirements for airlocks that a low both doors of an airlock to be open during CORE ALTERATIONS and ing movement of irradiated fuel inside containment provided tha administrative controls are in place to quickly close one door and es blish containment closure. The isolation valve, or t porary closure device, serves to limit the consequences of acciden The proposed change would ensure the isolation valves, or unctional equivalent, will perform their required containment closur function and will serve to limit the consequences of a fuel handling cident as described in the Safety Analysis Report such ) that the resul of the analyses in the Safety Analysis Report remain bounding. I considering the consequences of a 6 sign basis fuel handling cidents inside containment, the assumptions in the analysis f ake no redit for the containment as a barrier to prevent the t postu ted release of radioactivity. For events that would occur during COR ALTERATIONS or movement of irradiated fuel assemblies, containment c sure is considered a defense-in-depth boundary to prevent uncontrolled release of radioactivity. t. INSERT 3A-Sb 0 3.9-8 The proposed change to 18 months would apply the same frequency of testing to containment purge isolation valves as is applied to other containment isolation valves that must be OPERABLE during reactor operations. The 18-month frequency has been found adequate for the type of testing applied to instre.nentation and valves that must mitigate events much more severe and much more challenging to the containment boundary (e.g., LOCA. HSLB) than the FHA. l l

l .l i CONVERSION COMPARISON TABLE - CURRENT TS 3/4.9 Page 3 of 9 f a TECH SPEC CHANGE APPLICABILITY l i NUPSER. DESCRIPTION DIABLO CANYON CONANCE PEAK WOLF CREEK CALLAWAY 4-02 Removes SR to perform verification within 100 hours prior to Yes Yes Yes Yes i LS-4 the start of core alteration or movement of irradiated fuel. l 4 03 The frequency of verifying that [ containment ventilation Yes Yes Yes Yes LS-5 isolation] occurs is changed from 7 days to IS months. 4 04 Revised Surveillance requirement to allow for increased Yes Yes Yes Yes TR-1 flexibility in using an actual or simulated actuation signal. Identification of the specific actuation signal is j moved to the Bases. } r i 4-05 Moves SR to verify the trip set point concentration value No not in CTS No - not in CTS No - not in CTS Yes - Noved to FSAR LG for the Containment Purge Monitors is reset during CORE [ + ALTERATIONS or otter movement of irradiated fuel in contairment. 4-06 The requirement to verify the capability to close the No - not in CTS Yes No - not in CTS No - not in CTS ~ LS-23 containment ventilation isolation valves from the control room would be deleted. 4-07 The specific administrative controls used to assure No - not in CTS Yes No - Amendment 95 Yes LG personnel airlock closure capability would be moved from the did not put admin f LCO to the Bases. controls in LCO. 4 08 Moves the references to Heavy Loads in the Applicability and Yes No - not in CTS No - not in CTS No - not in CTS LG Action sections of CTS 3.9.4 (Contairment penetrations) to the Heavy Loads Program. 4-09 LCO 3.9.4 would be modified to permit an approved functional Yes Yes No - already in CTS Yes j LS-14 equivalent of a valve or blind flange to isolate containment (Amendment 74) penetrations. { 4-10 / a' f no s i ti ths t 4se-t#A yet AJ/A alfs M/A & 4/A f LS-20 vi dir t ces he at / i f tro. Not ab ~ - ^ ^ ~ WCGS-Conversion Coneparison Table-CTS 3M.9 S/15A>7

NO SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) CONTENTS .I I. Organization 2 II. Description of NSHC Evaluations 3 III. Generic No Significant Hazards Considerations "A" Administrative Changes..................... 5 "R" Relocated Technical Specifications............... 7 "LG" Less Restrictive (Moving Information Out of the Technical Specifications) 10 "H" - More Restrictive Requirements 12 e IV. Specific No Significant Hazards Considerations "LS" LS 1................................. 15 LS 2................................. 17 LS 3................ 19 LS 4................. 22 LS 5................................. 24 w ) LS6................................. 26 LS 7................................. 28 LS 8................................. 30 LS 9................................. 32 LS 10 LS 11 .............. @ thc4. -[34 iqs.v.11.t-A ..............................m 1 LS 12 38 LS 13 40 LS 14 Not applicable LS 15. Not applicable LS 16. 42 LS 17 Not used LS 18 Not applicable LS 19 t applji.cjlble . TAN 4.} p,3 g. ; j LS 20 LS 21 47 LS 22 50 LS 23 Not applicable LS 24 52 LS 25 54 LS 26 56 V. Recurring No Significant Hazards Considerations "TR" r TR 1. 58 WCGS-NSHCs-CTS 3M.9 1 S/2S/97

l v.iete. W a.9 - A IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 20 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE i REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS A note is added to LC0 3.9.4.c and the 7 day SR to state that contai t penetrations that provide direct access from the containment atmosp e to the outside atmosphere may be open under administrative controls. The te would allow these penetrations to be unisolated during CORE ALTERATIONS and ement of irradiated fuel assemblies within containment provided that s fied administrative i controls were employed. This note is in accordance with trav er WOG-76 to NUREG-l 1431. Rev. 1. The proposed Note is acceptable based on ahi strative controls that l l consist of written procedures that require designated pers nel having knowledge of l the open status of the valves in question and specified rsons designated and readily available to isolate the open penetration in t event of a fuel handling accident. These administrative controls provide prot ion equivalent to that l afforded by the administrative controls used to esta ish containment closure for a i containment personnel airlock. The NRC staff has a owed changes to the l requirements for airlocks that allow both doors o an airlock to be open during CORE l ALTERATIONS and during movement of irradiated f inside containment provided that l a&inistrative controls are in place to quick 1 close one door and establish containment closure. This proposed TS change has been evaluat and it has been determined that it i involves no significant hazards consider tion. This determination has been performed in accordance with the crit ia set forth in 10 CFR 50.92(c)as quoted below: l "The Comission may make final determination, pursuant to the procedures in 50.91. that a proposed ndnent to an operating license for a facility licensed under 50.21 ( ) or 50.22 or for a testing facility involves no signifIcant hazards onsideration, if operation of the faci 1ity in accordance l with the proposed ndnent would not: 1. Involve significant increase in the probability or consequences of an accide previously evaluated; or 2. Cre e the possibility of a new or different kind of accident from any a ident previously evaluated; or 3. Involve a significant reduction in a margin of safety. " l l The foi owing evaluation is provided for the three categories of the significant i hazar consideration standards: 1 Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? WCGS-NSHCs-CTS 3M.9 44 5/1587

{ 43A-7 l IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-20 (continued) The proposed change involves changes to the Technical Specification requir ts for l containment closure which is an accident mitigating feature. The changes w uld not l affect the likelihood of occurrence of any accidents previously evaluated The proposed change does not involve any hardware or plant design changes. he containment leakage value is not assumed to be an initiator of any an yzed event. The isolation valve, or temporary closure device, serves to limit t consequences i of accidents. The proposed change would ensure the isolation valv, or functional equivalent, will perform their required containment closure func on and will serve to limit the consequences of a fuel handling accident as descri in the USAR such that the results of the analyses in the [USAR] remain boundin. In considering the consequences of a design basis fuel handling accidents insi containment, the assumptions in the analysis take no credit for the contai nt as a barrier to prevent the postulated release of radioactivity. For ev ts that would occur during CORE ALTERATIONS or movement of irradiated fuel assemb es, containment closure is considered a defense in depth boundary to prevent unc ntrolled release of radioactivity. Additionally, the proposed change d s not impose any new safety analyses limits or alter the plant's ability to d ect and mitigate events. Therefore, this change does not involve a signif' cant increase in the probability or consequences of an accident previously evaluat 2. Does the change create the possibili of a new or different kind of accident from any accident previously evalu ed? The proposed change involves reliance n manual actuation of containment penetration valves or closure devices rather th automatic systems or passive devices (blank flanges or closed valves) to block he unimpeded flow of the containment atmosphere to the environs. The proposed c nge would not necessitate a physical alteration of the plant features that provid core cooling or subcriticality (no new or different type of equipment will be in alled) or changes in parameters governing plant operation during CORE ALTE IONS or movement of irradiated fuel in containment. Thus, this change does no create the possibility of a new or different kind of accid >nt from any accid t previously evaluated. 3. Does this cha e involve a significant reduction in a margin of safety? l The proposed cha e is similar to the use of administrative controls to isolate an l open containmen airlock door. The use of administrative controls in this manner l has been appr ed by the NRC staff for plant operations that would not require the containment o maintain a pressure boundary. This scenario is applicable during plant shu own for refueling when CORE ALTERATIONS and movement of irradiated fuel assembli s in the containment occur. The accidental damage to the spent fuel during these eration are classified as fuel handling accidents. The proposed change has been eveloped considering the importance of the containment boundary in limiting the/ consequences of a design basis fuel handling accident. The proposed change allows for protection equivalent to that provided by previously approved methods of WCGS-NSHCs-CTS 3N.9 45 5/15/97

l l h IV. SPECIFIC NO SIGNIFICANT HAZARDS C0 ERATIONS NSHC LS 20 l (continued) containment closure. Considering the proba ity of an event that would challenge the containment boundary, the alternativ protection provided by this change, and the operational requirements to occas nally open these penetrations, the proposed change is acceptable and any reduc n in the margin of safety insignificant. NO SIGNIFI HAZARDS CONSIDERATION DETERMINATION Based on the above eval tion, it is concluded that the activities associated with NSMC "LS 20" resulti from the conversion to the improved TS format satisfy the no significant hazar consideration standards of 10 CFR 50.92(c): and accordingly, a no significant azards consideration finding is justified. l I WCGS-NSHCs-CTS 3M.9 46 5/15197 l

l i l l i Industry Travelers Applicable to Section 3.9 _ ca a nd coa.;.tM th event pl>t of4fah8b. TRAVELER # STATUS DIFFERENCE # COMMENTS l l TSTF-20 Incorporated 3.9-2 Approved by NRC. (TST'31$Aar.1(([ncor F NA ^;; x u f 5,y 70 0. MTR. 3.1-001 TSTF-23, Rev. Incorporated 3.9-13 Traveler' bracketed ITS i 3.9.2 and revised the Bases for3.9.3. Bracketed Bases ) information from the traveler that is not W lay applicable to a specific plant g g(_,, was not incorporatedy lQ 3.cl-3 ) w TSTF-51 Not incorporated NA Requires plant-specific remnalysis 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. I Not incorporated NA Not NRC approved as of traveler cutoff date. TSTF-% Incorporated 3.9-4 Approved by NRC. MbNi)t 'Te s.9-a>2]. TSTF-136 Incorporated NA 3 TSTF-153 Not incorporated NA N.;IC C -ry. M M) T8Macorporated (.k.dM p3.1 r3 (~.usew) inco,,o,ated 3.,-1s ]c? 3.'t -I a.] $5al changes +o 33%. (notincorporatecL / MTR 5.1-062. ] l S/158 7

Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LC0 3.9.4 The containment penetrations shall be in the following status: The equipment hatch closed and held in place by f_oGr N@@N$ a. bolts; b. One door in eeeh tlkEasithnogy air lock closed andligst g.. _ m m. and

SW-immuz-2=zit wmuc--w?Mw~~~~~M A

NE 3

2iassGImaGa[E~ddle7nmHghantmfas'emeraergy; s$95'iBs'

~ . + a... 5 :. :,. MieV M M EZW98FM A-$ -gg[fEtM 3632*"j5 #f. J c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by an OPERABLE Containment EM Purge ; d ".;;st Isolation System aT W. gig;g d .....w.:.w ;..a. .w.v._ -s.. ....a...p.3.9.l{ y y Mb hk [ M ~'. I .l\\h e.l, j % '~ ? T,~ M & I K f : ~ ~ ..,3~ .v n s p. - p g..... m y y. y.-............. _ % Ann P-63 (Sey vic.a. Ah valw.s KA V-039 w A. VA y. II6) and Pencivath 9-98 (Bro}hh % valws KB V ool %Ks V cD2.} APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. WCGS. Mark-up ofNUREG-1431. ITS3.9 3.9 5 5/15/97

Containment Penetrations 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS. required status. ANQ A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is 7 days in the required status excep_t for__ containment E3i9E110l pghetrations.that are open under administrative g33,q ) c6ntrols. -gg g SR 3.9.4.2 Verify each required containment purge end 18 months ?PS2

2;u;t isolation valve actuates to the m

isolation position on an actual or simulated <

8,6 actuation signal.

WCGS-Mark-up ofNUREG-1431 -ITS 3.9 3.9 6 5/15/97

ws KA A awi, KA y-)2 (Gwtndrn ?'Penetratian 9 OhL Bre&*n3 Aiv v4Ns, KB V-coi %A.KB V 002. (Containment Pene.tration P-9ed mg b<. Containment Penetrations opened. urnA.an. admimstr.=b W conivo) d.wum3 roovemest B 3.9.4 W-radaa.ted. &cl er' CZXREAL.mE32ATLOQS 16 povYae, air sef v Us b.) BASESQeacer-bpid% b sqpar. cu.x49_ oz. Ah.ah.(Mef.s). 9Ml p BACKGROUND when containment closure is not required, the door interlock (continued) mechanism may be disabled, allowing both doors of an air lock to i remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiated fuel 3 assemblies within containment, containment closure is required; l therefore. the door interlock nechanism may remain disabled, b;t er.e eir leck deer ;;;at elay; rair, cicxd.Qj@ igg,3ne3iipiaggiik.ai,r MliibrMla_gijbTfstE4igBDMRitC1l lie ~Glie~616ck i d!I! DIE!EBILWa!$L~]h2heJREBEREiesen!MBR3tt!!IOh!O tueteesmuseneimmyr ga g g ag e n - - w-lock e aRMath g C 4Lh T R-T T __ _ JJ_C22it';the i t c= -Mesic-iwodaisrmistananz_. _ -r Igess!gs .ansM w w I=-1~f4-nIniWidhetW 2 N 74 Lasioif IEENWM ANEBENRRWlEspec1gniinMaitnHLIGRenRMMgrande j ,qDalliiWit$ade;the WIToE2DWNfanliti7amitlMaatMDet3imitiid i WWfdklBEnMMEnfsisflhI !ileU3EBiniEMMMietions - W __:R: MR 2roshihFtW8V~tlWeh aNfB[otf,umuMisiidTir.~ ; C W ~ k the., servict. dtv" odbreMbi_ng Qif VAIV3t 0" The requirements for containment penetration closure ensure that a relecse of fission product radioactivity within containment will be l restricted from escaping to the environment. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling. The Containment Purge ;r.d Oimt System includes two subsystems. The normet Stiutdoiln" Pin:ge subsystem includes a 48 36 inch purge sigigy penetration and a 42 36 inch exhaust penetration. The second subsystem, a minipurge system, includes an 8118Jinch purge s@]y penetration and an 818 inch exhaust penetration. During H0 DES 1, 2. 3, and 4, the two valves in each of the norme+ shutdown purge ly l SvWby-and exhaust penetrations are secured in the closed positiongr, hbildd- "Q. The two valves in each of the two minipurge penetrations can be opened intermittently, but are closed automatically by the [ "D Engineered Safety Features Actuation System (ESFAS). Neither of the j subsystems is subject to a Specification in H0DE 5:orJE@Ei6]pxcluding l W AL'I RAT DIS _p tuo.vementioL jrradiated_fue]51n? containment. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.9 B 3.9 13 S/1587 t

j Containment Penetrations B 3.9.4 l l BACKGROUND In H0DE 6, large air exch.; ages are necessary to conduct refueling j (continued) operations. The normal 42 36 inch purge system is used for this i purpose, and all four valveseMDBie closed by the ESFA3 in accordance with LCO 3.3.26. "6TIMEttD4plition C.gira; red 0;f;t, Tutur; ^,ctati;n Sy;t;; 'EST'0) Instrumentation-l" [ NNW1MMWDTJgiggg!sisLment_. Ei ;iGiFi y. ;y;t;; i ~ iG'; eiar;tieral iG E 0. ;r.d ;ll f;;r ......,m. .... ~,..,,. ~ -. s i 1 i Iggn the minipurge system is not used in H0DE 6 ',--AR aR four j 18 inch valves ar; x;;r;d in tt.; closed p;iti;n. The enher containment penetrations that provide direct access from containment ateo re to outside atmosphere must be isolated on atW-7 \\ j least one si Isolation may be achieved by an OPERABLE automatic 1 atfon valveTor by a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations giCtturteenimacy 6'I]RKiduring fuel movements (Ref.1). j APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY ANALYSES assemblies within containment, the most severe radiological 4 consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in Reference 3 2. +nehde adsomes dropping a single irradiated fuel assembly ;ad Par.dlir6 t;;l

; t; ;, ;b1;;t.at; etter irr;di;ted f=1 n d li n.

The l requirements of LCO 3.9.7, " Refueling GevMy M Water Level," and the miniaun decay time of 100 hours prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the 3 guideline values specified in 10 CFR 100. Standard Review Plan, i Section 15.7.4, Rev.1 (Ref. 3), defines "well within" 10 CFR 100 to i be 25% or less of the 10 CFR 100 values. The acceptance limits for { offsite radiation exposure will be 25% of 10 CFR 100 values ;r-tra =C 1 (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.9 B 3.9 14 S/15 8 7

Containment Penetrations i B 3.9.4 l BASES } d 4 APPLICABLE staff sppr;V;d licasing basis 'c.g. a spaified fractica of SAFETY ANALYSES 10 C",100 li;;;it ;). l (continued) i Containment penetrations satisfy Criterion 3 of the N",0 "clicy i Stat;nat 10CFR;50:36(c)(2X...ij).. j 1 --~

-. _ _ _ _. e=rgec-c_

.----.w-- -. ~ _m--__ ,,m_.

s. t.m_ _.-. - --

-, - - ~ _,_mm=--, - - - - -. =,., n= _1 = =--.,m_ ___,....m_.---_-- . m. 2 a- =. q w._ . e w q___samy _.--..----_--;ams=u-1 ars. = _ n;._ - - _ s. e- .n-__,_ m_= -n,-_ -xw s___ s..--, w s --c----1.m_. _n..__,..,,_-~,--:-- = i ,,, - - =.. .w -- m- ) _.._.m_ - m u._w-- -- _.m--._ - m.....-m---- ...w - - wm-= sm. a :__,. _..m - -- - mm w,.,_ n - w m.u.__. f.. a.s _..v_-_ =. . _-_,, 3.m. ..--- -,m.,,_:......_,. m,, m. ~ = _ - ..w.-__.. - -. - = =, _ - - -- l .A _ m 2 - - 2 wah-

w-*~cm ann

=- -g; w _-.- w -s - -1 v_,m -m _ _.c=- .,.m..._-*=+==.-=_--._r==une-r.- .,,.__z _x__- .,,, ~ ...m-cmu.m -s .- ~ _ - - - l LC0 This LC0 limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any j penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge ;nd d;ust penetrationsstodEthe)ersonnat iaitlock. l For the OPERABLE containment purge ;nd daust penetrations, this LCO i ensures that-thesesach penetrations-ere B isolable by the Containment Purge and [2;u';t Isolation Systc;;;. The 0"C",.'."_a!LI"' 4 3 ,,.m.._m_,__m....._. m...__. u...m. u.. m, m..- ,,.m ...i___. _. _ m _m y .r. m ~ -~ m m Vrs1 VC C C ^.u rd III.C ^, ^.hC Ci fi Cd i n I v6 Ci.I. bC C siCYC An. ther;fer;. x;t the es';,...ptions u';cd in tri :;afety s=1ysis to ensure i that releases through the valves are terminated, such that l radiological doses are within the acceptance limit. i @0I3!9743_b~Km. j[difigbEa'NotC61Tdiiritiglad dii[ergenci]escapelait 1 Hic.k.itt.elp.prafBjosure:d.evic.e: tot.b.e.' arf.aic.t:ep.tabitFreplacement:for an emitrggnrair~diurtfoori 4 (Q30 i l (Ttgg _1__n ::Mi _

- %d-.;__f.: !tgMaWwich)

Le0~3 sxmmetradwarmtmawirisDLeuccesen 3 _ -._ r 1 i

MManisolated tutderrathitemweinTcentrtils? -

h#Md i 1 . athatt c. ms 1-awa st a df r- ~ ~ on.. 1 -. 4. .T... u ..f L [the' h7t6e~ lare. _ gna . _.11F_ il eit6_ sol _.._ l._ _ t. ~ Jandr ( f~aWeT' hand 11ngr dent R.netrahm P-63 (.seniu. & va%.s. KA v-@ ud KA V-tied u (continued) i ene.tmbdn P.9B(Bre2&Mg AM valve.s KB V. Col 9.d. K B V-co2') w WCGS-Mark-up ofNUREG-1431-Bases 3.9 8 3.9 15 5/15/97 p

Containment Penetrations l B 3.9.4 i BASES l APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In H0 DES 1, 2, 3, and 4 containment penetration requirements are addressed by LCO 3.6.1. In H0 DES 5 and 6, when CORE l ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling l accident does not exist. Therefore, under these conditions no l requirements are placed on containment penetration status. l ACTIONS A.1 and A.2 If the containment equipment hatch, air locks. or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge ;rd ".xh;u;t Isolation not capable of automatic actuatior J.ca the purge erd exh;u;t V;lv;; ;re open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by imediately l suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a e.omonnent to a safe position. SURVEILLANCE SR 3.9.4.1 REQUIREMENTS l This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open purgc ;rd exh;ust m on valves will demonstrate that the valves are not blocked from closing. l i Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being I closed by an OPERABLE automatic containment purge c.rd ;xh;;;t isolation signal.1 7. ". w r w o+ ew n v'. v 4, .-j.--r .mm, .m ..-..._.m__ .. ww - dR d@1rfttgrtim3trpenetNt1&nifDtilE gig p,q., j l The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. The (continued) WCGS-Mark-up ofNUREG-H31-Bases 3.9 8 3.9 16 S/lS/97 l

a j Containment Penetrations 1 B 3.9.4 i I .t ) BASES (continued) 1 i ' SURVEILLANCE SR 3.9.4.1 (continued) j REQUIREMENTS Surveillance interval is selected to be commensurate with the normal duration of time to compNte fuel handling operations. 4 surveillance before the start of refueling operations will provid; tw ;r tt.ra i WNcigg surveillance verificatione during the applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel ) handling accident that releases fission product radioactivity within the containment will not result in a release of fission product radioactivity to tt.; =vire.. ;., trit 2WI6DRBli2Nefie. l SR 3.9.4.2 This Surveillance demons'trates that each containment purge 11i2Tintion

rd ; feat valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The l

18 month Frequency maintains consistency with other similar ESFAS i instrumentation and valve testing requirements. In LCO 3.3.6, the Containment Purge emHbeheest Isolation instrumentation requires a CHANNEL CECK every 12 hours and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 18 months a CHANNEL CALIBRATION is performed. ";; ;y;ta ;;tstim rug,r.x ti; i; i _.ati;ted ;nry 10 r.e;;tta,171r6 r;falirs, = ;.7, ffdC "i BAES: SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surve111ances grferad irir; = 0 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment. REFERENCES 1. GPlHiuele;r 0;f;t3 Enlatier, 0 0000000 001, n;v. O, Foi 00, 1000. Almen4prgg316~1ieritirattfjgiStation i Gens _tsgtLL%enseE42Ldltg(MEMI 1 2. FSAR USilR, Section 153 3 (15.4.0) 3. NUREG 0800, Section 15.7.4, Rev.1, July 1981. 4I AMladskritX95]tojWolfECreekJianerating; Station _ Operating f5.- LWWWZdated.hryl28D996.' g,q,q j ._n StsNcm[- Amerdmed 40. (D l b WMGeek Gemerst,m4 \\ CPecth beamsa. MW-42., McA. Jg ti, t'57. WCGS-Mark-gp ofNUREG-1431-Bases 3.9 8 3.9 17 $/15M7

.. - ~. _ _.. _ i-3 i l i CHANGE NGBER DESCRIPTION li 3.9 6 A note is added to LCO ".9.4 stating that an emergency personnel ) I escape air lock temporary closure device is an acceptable i j replacement for an emergency airlock door. Since the temporary closure device for the containment emergency personnel escape air lock will be fabricated to the requirements for use during CORE [ ALTERATIONS or movement of irradiated fuel within containment, j and utilized installed, and tested in accordance with plant procedures, the temporary closure device can be expected to perform in a manner equivalent to that of the emergency personnel j escape air lock during a design basis event. This change is } consistent with the current TS. l j 3.9 7 In accordance with current TS, LCO 3.9.4 status item c.2, and j associated Bases, would be modified to state " capable of being closed by an OPERABLE [ Containment Purge Iso'ation Valve]" (i.e., { the word valve would replace the word system.'. I 3.9 8 Not applicable to WCGS. See Conversion Comparison Table j (Enclosure 68). i l 3.9 9 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). M3M-( l r -w 3.9 10 Consis it c ,I .9 icab yi bei i rev te " r C , exc dur i i t of tr - rive fts abi y1 ing is e th requ is li ive a re at ech al s ific on r uir t t yf reac vess wat av duri mov of tro rods j i [(re ated nic s ica n 3.9 .2)] The ocat ific on p es oe:r mit o he nt of ater a ve the t of fue ssemb es in rea or ve el duri 4 ( t to 01 rods. The ses S 3/. 10)] te t fthis sure water ramo 99% the used 10 iodi gap ivit eleased'from t ruptur of a tradiat fue ss y in t event of a f and11 acci t (FHA). r, t emen contr rods i ta 1ated t initi er[ {I ,condi ns of ,a B s do no ddre any r rding dvertan ritic ity whic could ead t a bre6ch of the f rod cla lH$EC 6A-Za 3.9 11 I da witAt tr ele L is i ed i 7 a net io ath a rovi sd ect c s fr d n nme to o 1 a her o un ola Q ini _ at co rol. T all nce Ao h e ta fJbw WCGS-Differencesfrom NUREG-1431-ITS 3.9 2 S/1S/97 l

CHANGE NUEER DESCRIPTION lg3,q,9 } hthe ith irec cce [ the ntai ta spher to t ] ide re sol ed r ahi stra e rols s on fi ory se c lati of a uel ndli i whi indi ea table iol cal s es nd to 1 ada stra ve cont st ens tha he i tr ons 1 be tly osed 1 af ha 1A l ace provi a def e in-th a roach o / l cept e dose onseg es. ahi strat ec rol ( re def in t Bases. 3.9 12 A MODE or other specified condition of applicability change } restriction has been added to ITS 3.9.6 in the LCO Applicability per the matrix discussed in CN 102 LS 1 of the 3.0 package. This restriction prevents reducing the water level above the reactor vessel flange while the RR function is degraded. (See l the LS-1 NSHC in the CTS Section 3/4.0, ITS Section 3.0 package). 3.9 13 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). i l 3.9 14 A MODE change restriction has been added to ITS 3.9.1 in the LCO i Applicability per the matrix discussed in CN 102 LS 1 of the 3.0 package. This restriction prevents a transition from NCDE 5 to { MODE 6 if the boron concentration limit for MODE 6 is not met. (See the LS 1_NSHC_in the CTS Section 3/4.0, ITS Section 3.0 package). @ireRT p % o3,4 m g q,,,.3, y l 3.9 15 LCO 3.9.1 has been revised in accordance with traveleMMto l clarify that boron concentration limits do not apply to the 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 suberitical 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 i boron concentration in these areas when they are not connected to the RCS. This change is consistent with the intent of the Specification, as described in the Bases, and eliminates restrictions that have no effect on safety. 3 A-lb lusEKT Gk -Sb {Q3A I } I WCGS-Diferencesfrom NUREG-1431-ITS 3.9 3 S/1S/97

. _. -._..- ~. _ m.__. l l INSERT 6A-3b 0 3.9-7 I' 3.9-16 ITS LC0 3.9.4 is revised to add a Note consistent with CTS 3.9.4. The Note will allow Penetration P-63 (Service Air valves KA V-039 and KA V-118).and Penetration P-98 (Breathing Air valves KB V-001 and KB V-002) to be unisolated under-l administrative controls. Amendment No. 107, dated July 11, l 1997, revised CTS 3/4.9.4 to add a new limiting condition of l-operation to state that Penetration P-63 and Penetration P-98 may be opened under administrative control during core alterations and movement of irradiated fuel assemblies within i -the containment. Additionally, SR 3.9.4.1 has been revised to l except containment penetration P-63 and P-98 if they are open l under administrative controls to comply with the ITS format and rules or usage-(e.g., SR 3.6.3.3 and associated Bases). I I i I l. f. I f' k l

\\ t CONVERSION COMPARISCN TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 3.9 rages era DIFFERENCE FROM NUREG-1431 APPLICABILITY NUISER DESCRIPTION DIABLO CANYON COMANCE PEAK WOLF CREEK CALL M t 3.9 15 LCO 3.9.1 has been revised to clarify that boron Yes Yes Yes Yes concentration limits do not apply to flooded areas that are not connected to the RCS. N S E R.T 6 8 - 3 l C 3.9 I } l I i WCGS-Conversion Comparison Table-ITS3.9 5/1SAF7

INSERT 6B-3a 0 3.9-7 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.9-16 Wolf Creek ITS LC0 3.9.4 is revised to add No No Yes No a Note consistent with CTS 3.9.4. The Note will allow Penetration P-63 (Service Air valves KA V-039 and KA V-118) and Penetration P-98 (Breathing Air valves KB V-001 and KB V-002) to be unisolated under administrative controls. Additionally. SR 3.9.4.1 has been revised to except containment penetration P-63 and P-98 if they are open under administrative controls

. ~ . MTIONAL INFORMATION COVER SHEET ADDITIONAL INFORMA

MO: O 5.5-4 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 5.5.4 e and Difference 5.5-13 Comment: WOG-72 has not yet become a TSTF. Use current ITS.

FLOG RESPONSE: This change to ITS 5.5.4 e was prepared in accordance with WOG-72, Rev.1 which is currently under TSTF review. The change specifies that the requirement to determine cumulative dose contributions from radioactive effluents need be done on a current quarterly and annual j basis instead of every 31 days. We believe there is a strong technical basis for this change to the ITS We request that the NRC keep this as an open item under the assumption that the traveler will be approved prior to issuance of the SER. FLOG RESPONSE: (revised) Based on the present status to the generic change process for the STS, it appears that traveler TSTF-308 (previously WOG-72) will not be approved by the NRC in time to support the initial license amendments for the FLOG plants. In order to facilitate the issuance of these initial license amendments, an alterriate approach has been developed which relies on the CTS, plant specific information and/or the NUREG but does l not rely upon the traveler. This altemate approach is hereby provided as an interim submittal to allow issuance of the initial license amendments. The changes which rely upon the traveler can be processed in j subsequent license amendments following approval of the traveler by the NRC. ATTACHED PAGES: i Attachment No.18, CTS 6.0 -ITS 5.0 Encl. 2 6-17 Encl. 3A 4 Encl. 3B 3 Encl. 5A Traveler Status page,5.0-10 Encl.6A 3 Encl. 6B 4 l l i l

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

e. Radcactive Emuent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactrve effluents and for maintaining the doses to MEMBERS Of THE PUBLIC from radioactive emuents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actons to be taken whenever the program liraits are exceeded. The program shallinclude the following elements:

1) Limitations on the(functional-: ::' "Y;: ::S; of radioactive liquid and

~0205-XT gaseous monitoring instrumentabon including surveillance tests and 2DM astpomt determmation in accordance with methodology in the ODCM,

2) Limitations on the concentrations of redcochve material released '

in liquid effluents to UNRESTRICTED AREAS conforming to C 92-14-M ' (10 times the concentration valuee in110 CFR "W ' - { Part 20, Appendix B, TatWeQJih Column 2,(to 10 CFR 20.1001-20.2402.]

3) Mondonng, samping, and analysis of radioactive liqub and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, I

4) Lmtabons on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioachve materials in liquid emuents released to UNRESTRICTED AREAS ce&..as to Appendix 1 to 10 CFR Part 50, 3

re-. x

5) Determmation of cumulative ::d ;::,: ':Idose contributions frorn radcactive effluents for the current calendar quarter anjd

' M-cunent calendar year in accordance with the and @ 5.9-+ \\ n_ e in the ODCd '- " : : ; ?' _:; _E_^^_. 5--1;o n e s

6) Limitations on the[ functional capability' ;:. 217 */ and use of the liquid

~ 0245 X~T and gaseous effluent treatment systems to ensure that the appropriate

4C portons of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day penod would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix 1 to 10 CFR Part 50.

C4C LthaA M M 8

7) Umitatens of the dose rate og ioactive material

% El-I \\ released in gaseous emue ., - <n,.w,, SITE BOUNDARY area ond the q u e-- a w s---. _,.,n _, T: ': F, bb( di tE,5EsAArtAM

Add

w. ii. -

f a. For noble gases: Less than or equal to a dose rate of 500 mromlyoar to the whole, body and less than or equal to a dose rate of 3000 mrom/ year to the skin, and

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

{ mrrn/ year to any organ. WOLF CREEK - UNIT 1 6-17 Amendment No. 42 Mark-up ofC156.0 $/1587

CHANGE NUMBER HSBC DESCRIPTION i (referenced in the ITS or CTS). covered under the provisions of ITS 5.4.1.e. or required by 10 CFR 50, l Appendix E and 10 CFR 50.54(p) and (q). Therefore, these l requirements are duplicative and unnecessary. l 02 04 LG The In-Plant Radiation Monitoring Program is based on NUREG 0737 Item III.D.3.3, " Improved Inplant Iodine Instraentation Under Accident Conditions." The CTS l requirement for this program is deleted but the program i commitment will be in the USAR and implementing procedures. Any changes to the programatic provisions in the USAR would be controlled by 10 CFR 50.59. Moving this l information maintains consistency with NUREG 1431. Rev. 1. 02 05 A Revises the radioactive effluent controls program to l reflect wording in NUREG 1431 and format of the Administrative Controls section. The term " operability" is replaced with " functional capability" to avoid l confusion with the TS definition of operability. 02 06 A Consistent with NUREG 1431, the Radiological Environmental Monitoring Program is deleted. The details and description of the program are duplicative of [0ffsite i 3-Dose Calculation Manual (ODCM)] requirements. The program l only maintains consistency with the requirements of 10 CFR i 50, Appendix I. The [00CM] and regulatory requirements i provide sufficient control of these provisions, and therefore, removing them from the CTS is acceptable. l l 02 07 Revises dio ive Ef nt rols ran se ! proj ion o meet igin inten f TS p r to yQ~ ipp(eme tion L8 1. Gl. 01 pr ided t wording dor STS tio .5.4. which ined

r uir s fo umulat and ject se.

s requi ap t to e pro ed do for quar fa year a 31 basi. It i nly ssary reaso e to eap jectio or t xt 3 ays. ative se pro tion still equir or t i l rrent lender uarte nd ye in acc ance y h Jhe L _ This __- nge i consist t__with raveler 10G-72. 02 08 H Revises the procedures section to refer to all programs listed in the program section of the Administrative i Controls and ensures implementing procedure control for these programs. This change is consistent with NUREG-1431. This change is more restrictive because it now includes all programs (current and added) in the TS as well as those already specified in this paragraph. WCGS-Description of Changes to CTS 6.0 4 5/15/97 l l

s CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 3 of 8 TECH SPEC CHANGE APPLICABILITY C#EER DESCRIPTION DIABLO CANYON CONANCHE PEAK WOLF CREEK CALUWAY l t 02-01 CTS Section [6.6.la] for Reportable Event actions has been Yes No. Deleted from Yes Yes A deleted from the CTS. CTS per Amendment l 50/36 l 02-02 The " Safety Limit Violation" section is deleted from the Yes Yes Yes Yes LS-4 CTS. The NRC reporting requirements are duplicative of 10 CFR 50.72. 10 CFR 50.73 and 10 CFR 50.36. 02-03 The iglementation procedure requirements related to [the Yes No. Deleted from Yes Yes A security plan, the emergency plan.] process control CTS per Amendment programs and radiological environmental and offsite dose 50/36 calculation programs are deleted from the CTS. 02-04 The In Plant Radiation Monitoring Program is deleted from Yes. Move to FSAR. No. Deleted from Yes. Move to USAR. Yes. Move to FSAR. I LG the CTS but the program cannitment will be in the USAR and CTS per Amendment iglementing procedures. Any changes to the programmatic 50/36 provisions in the USAR would be controlled by 10 CFR 50.59. 02-05 Revises the radioactive effluent controls program to Yes Yes Yes Yes A reflect wording in NLREG-1431 and format of the Administrative Controls section. The term " operability" is replaced with " functional capability" to avoid confusion with the TS definition of operability. j 02-06 The Radiological Environmental Monitoring Program is Yes No. Deleted from Yes Yes t' A deleted. The details and deaription of the program are CTS per Amendment duplicative of [00CM] requirements. The program only 50/36 maintains consistency with the requirements of 10 CFR 50. Appendix I. 02-07 'Re es ioa' iv ff t tjr ffs Pf gr h dh h WA h W/A h >J/A 6 A oj ons o to gin inMnt of tii g ti t i of 89-e4o+ lised. ~- WCGS-Conversion Comparison Table-CTS 3M.0 S/158 7

( INDUSTRY TRAVELERS APPLICABLE TO SECTION 5.0 TRAVELER # STATUS DIFFERENCE # COMMENTS i TSTF-9, Rev.1 Incorporated NRC approved. l TSTF-37, Rev.1 Incorporated 5.6-2 DCPP only. l TSTF-52 Incorporated 5.5-4 (["[fI[q'h[*Gha 5.0 -Gj l TSTF-65 @ Incorporated NA h_NRC appros ag.easj j Q s.1-9 Wesdifterd _ ) TSTF-196, Rev.1 Not Incorporated NA Retain CTS. TSTF-118 Incorporated 5.5-8 [pp r@lTeF.o-m(-I f @M1[ j[eflueprporatej - MA /'Retajn-CJ}8 [ rn c.o -ec. J TSTF-120h Not Incorporated NA Retain CTS Prn c.o-os 1 @M IModomfeV_ / 5a-O ia s.2-0 4 l TSTF-152 Incorporated 5.6-4 gQ 4 ra co-eQ -f (TFffA11Ek / W e g rated / / 5:7C f @ S'E-33

h( (V W Incorporated 5.6-5

@_ipprove.4)-tTR 5.0-uQl ) Q#6Gff/ Anssp9rsted' f Als)- Ns.s-4\\ M Incorporated 5.5-14 l* *'5-2 i (rmpMJrittaier Incorporated s.2-2, 5.5-1 s.2.3 upacgsn p, z., i s t:,, p.,=,, s.c->s, q rste -2 se e 4. i 4 S/158 7

Programs and Manuals 5.5 i 5.5 Programs and Manuals and setpoint determination in accordance with the methodology in i the 00CH: l b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming l to IEtggs;1%lelconcentration;valfmiEin 19-6FR-09, Appendix B, $$3&My$ Table 2, Column 2?[t'o:10;CFR?20?1001:20~.~2LO2: i l Monitoring, sampling, and analysis of radioactive' liquid and c. gaseous effluents in accordance with 10 CFR 20.1302 and with the 3 methodology and parameters in the ODCH: d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I: re m sMkeod e. Determination of ctmulative er.d i; ;j;;t dose cont ibutions from radioactive effluents for the current calend quarter and 4 current calendar year in accordance with the me logy and parameters in the ODCM[;t k;;t r;;rj 011.7. juster 1Rnet_ _.. g_q f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conformino lo_10 CFR 50, Appendix I: C b_ h sWe7 M g. Limitations on the dose rate \\resulting fromradioactivematerialNSA-'k released _in gaseous effluents *to areas'beyond the site boundary Wt; tt.: i;; ;;;;;;i "d.;itt.1^ C" 20. /GEr.dk C. _ n&5353U!3 T;;k 2. C;1= 1,fguttre sWn be

a.ccudrhu.T

% m.r uau : J n 1 FonnobleJases; 9.es(thsh ortouaPtp a dose rate $500 mres/trito.;thegwhole body; and Uh -wMfDa. dose - ratnM30_00 ares /yr toithefskinZand l 2. For,Iodinejl31;; for Iodine-133, ~for tritita. andlfor all M(515H1W l l (continued) l WCGS-Mark-up ofNUREG-1431-ITS S.O 5.0-10 S/1SM7 t I i

i CHANGE NLMIER JUSTIFICATION l 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies." to provide consistency with current application of these requirements. This is consistent with the use of current TS and l alleviates potential confusion in the program descriptions. This i change is consistent with traveler TSTF 118. il' 5.5 9 Not applicable to WCGS. See Conversion Comparison Table (Enclosure - 68). 1 5.5 10 Section 5.5.12c specifics a surveillance program to ensure that the quantity of radioactivity contained in all outside liquid radwaste l tanks that are not surrounded by liners, dikes,.or walls is less than l the predetennined quantities. This change lists the tanks that the surveillance program is applicable to as is in the current TS. This l change is a plant specific requirement consistent with the current TS. l 5.5 11 The docunents referenced for the testing frequency for the Ventilation l l Filter Testing Program (VTFP) do not provide frequencies for combined pressure drop tests or the heater power rating test. The current TS frequency is added for these two tests. l 5.5 12 Not applicable to WCGS. See Conversion Comparison Table (Enclosure los.s-47 5.5 13 (T angrev es R oacti Eff nt Co rols gram se (ons mee rigi int of T prior o imp 1 nta on fG 89 0 rovi the ording r the ( ion .5..e) Rek uxL c ombi the r uir ts for umula ea rojec se. r es a nt t ke p ected ses the art a yea l na day is. is on1 ess ya easona e mak a fp ecti for t next 31 ys. cumul ve do pr ecti i till requir for curr cale quart and y ri acc ance with l [ is cha is con stent ith tr eler 7-T5TF-23~7 i ITS Section 5.5.7 is being revised consistent with travelerh [and ( 5.5 14 ii ndmen+ No. Oserfte ametiftent M=stMitterD*1m]. The proposed i changes to ITS 5.5.7 provide an exception to the examination requirements in Regulatory Guide 1.14. Revision 1. " Reactor Coolant Puup Flywheel Integrity." The proposed exception to the l recommendations of Regulatory Position C.4.b would allow for an l acceptable inspection method of either an ultrasonic volumetric or surface examination. The acceptable inspection method would be conducted at approximately 10 year intervals. This change is consistent with the NRC Safety Evaluation Report associated with i WCAP-14535. " Topical Report on Reactor Coolant Ptap Flywheel Inspection Elimination." i i-WCGS-DufferencesfromNUREG-1431-ITS5.0 3 5/158 7

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 4 of 5 SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUtBER DESCRIDTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.5-12 The referenced frequencis for the tests listed in the Yes No No No Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCFP (see LAR %-09) oj is mee ori i 4 t on GL _-. (WOG-72). & M-f 19G S*4 I 5.5-14 Section 5.5.7 is being revsied consistent with traveler Yes Yes Yes Yes [and @MNJe@eMW Meeen Am,ngmQ4 pig toir, {q s.s-2-\\ ( The proposed changes to Section 5.5.7 provide gg 3 q gg an exception to the examination requirements in Regulatory Guide 1.14. Revision 1. " Reactor Coolant Pump Flywheel Integrity." 5.5-15 This change provides a time interval of within 31 days No No No Yes after removal in which a laboratory test of a sample obtained from the charcoal adsorber must be tested. This change is consistent with Callaway CTS. i 5.6-1 Revises Section 5.6.4. " Monthly Operating Report." to No. DCPP CTS Yes. LAR 9414 No. Wolf Creek CTS No. Callaway CTS reflect a revised submittal date. consistent with consistent with consistent with NUREG-1431. NUREG-1431. NUREG-1431. I i 5.6-2 Deletes the EDG Report to reflect the K-..dations of Yes No. Not in CTS No. Not in CTS. No. Not in CTS. equir f De ---{d? 5.6 ] ~ .dat y 31. TS TF-57, Rev i. _ 5.6-3 Revises report dates in ITS 5.6.2. " Annual Radiological Yes. Consistent Yes. See LA 42/28. Yes Yes Environmental Operating Report" to be consistent with with CTS and LA current TS. 78/77. [ssstr 66-4@} Q s.s ; 1 (S. Si - It. tusERT GS-A@l o s.z-i ) ~ s.s-let lo M RT. 6Eb - 4bpl cA S.o -oG ] @.5-Id tuseRT Gb -4c}.} t>c s.o -co3 l S/15/97 WCGS-Conversion Comparison Table-ITS5.0

l Att: chm;nt 2 to ET 98-0107 Pege 1 of 1 LIST OF UOMMITMENTS The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporation (WCNOC) in this document. Any other statements in this submittal are provided for l information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr. Michael J. Angus, Manager Licensing and Corrective Action at Wolf Creek Generating Station, (316) 364-4077. COMMITMENT Due Date/ Event A supplement to Reference 6 will be provided at a later date. Prior to issuance i of SER. j i l i l l l l A-}}

ML20198H105

Text

Reference:

Subject:

Navigation menu

Search