ML20154S591

Proposed Tech Specs Sections 3.1, Reactivity Control Sys, 3.2, Power Distribution Sys, 3.4, RCS, 3.5, ECCS & 5.0, Administrative Controls

ML20154S591
Person / Time
Site:	Wolf Creek
Issue date:	10/23/1998
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20154S589	List: ML20154S584 ML20154S591
References
NUDOCS 9810280029
Download: ML20154S591 (26)

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{{#Wiki_filter:... - ~ - - e l l REACTMTY CONTROL SYSTEMS l POSITION INDICATION SYSTEMS-OPERATING LIMITING CdNDITION FOR OPERATION t 3.1.3.2 The Digital Rod Posstion Indication System and the Demand Positiw Indication System shall be OPERABLE :nd ^ ;d': ef f: : "n, th: cent: r^d ggg{} pf' : ^:T::1. t2 : _;: APPLICABILITY: MODES 1 and 2@ Pgg g ACTION

a. With a maximum of one digital rod position indicator per inoper.

} able(for one or more groups)either:

1. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at least once per 8 hours and

'n:^"','within 4 hours)after any motion of the nonindicating rod gggp hungy l which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL PCWJR to less than 50% of RATED THERMAL POWER l

within 8 hours @ l [3. Be in HOT STANDBY within the next 6 hours.) Md@ man @n# wa

b. With more than one digital rod position indicator per benis(gEup)

{gg inoperable either: ) 1.a) Determine the position of the nonindicating rods indirectly i by the movable incore detectors at least once per 8 hours and @##"M N l

• x: _^M,LO;n 4 hourslefter any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, and

%} lQ3.l-193 b) P' - S entre! rM: =dr renut! tun!, end .l c) " nM: :nd ferd P r..Y C^^!:nt 9; tem :=r::: tempre:cre (T g Oc rt x x prheer, nd ~ ~ ~ h estore the digital rod position indicators to OPERABLE status @M in 24 hours such that a maximum of one digital rod position indicator per banis Ris inoperable, or I"13-0FA7 j was I

2. Be in HOT STANDBY within the next 6 hours.

1 l l l

• Separate Condition entry is allowed for each inoperable rod position indicatof

?%IMX5!d l and each demand position indicator.

1. 1. NM l

WOLF CREEK - UNIT 1 3/4 1-11 Amendment No. 44,89 9810200029 981023 PDR ADOCK 05000482 p PDR. Markup ofCTS3M.] S/ ism

i CHANGE 4 NL#EER NSlfC DESCRfPTION a Not applicable to WCGS. See Conversion Comparison Table 13 08 (Enclosure 38), g 1 ru use O_ ~-- s 13 09 ,pP23 Current TS ctions b.1. and b.1.c C0 3. 2 are i delet SOM is en ed in MOD and y rod tion. Mul le inopera DRPIs wi have impact o DM in S 1 and 2 the con rod ition ar verifi y 4 alternate ans and r motio s limit onsiste wip l f the ac nt analy 's, ion of t. e requir nts'is 5 YW Not appi cable to wuc. See converswn Comparison Table 14 01 (Enclosure 38). 15 01 o WCrK See GefiversioTGtfmpariserf Tabh]r Not icab e4.4 daad., nelosur __8 ). j Not applicable to WCGS. See Conversion Comparison Table 15 02 (Enclosure 3fi). j 16 01 LS 14 This TS would be revised to apply to shutdown " banks" ? instead of shutdown " rods"; this is consistent with NUREG-1431. Rev. 1. The current Action Statement permits one rod to be inserted beyond the limits: the proposed ITS l CONDITION A would allow one or more banks to be inserted { f beyond the limit. 16 02 M The proposed changes to the Action Statement would require that the shutdown banks be aligned within limits and that SOM be verified or restored. The new Action Statement would extend the time to achieve alignment from 1 to 2 hours as justified in the Bases for ITS 3.1.5. The new Action Statement would establish a Completion Time of 1 hour for verifying and restoring SOM. In the proposed Action Statement, both the realignment and the SOM verification would be required. The current Action ~ Statement provides a 1 hour limit to achieve realignment and effectively applies a 2 hour Completion Time to SOM verification and restoration (which would be performed ~ under the TS for rod group alignment limits). In the current Action Statement, either the realignment or the SDM verification are required. The current Action Statement could, in some circumstance, allow continued POWER OPERATION with a shutdown rod out of alignment because it was written to apply to individual rods and refers to the rod group alignment specification. The new action statement, which applies to shutdown banks, would not pennit operation with a shutdown bank outside its l WCGS-Description of Changes to CTS 3.1 13 S/25/97 i I ^

y@J.1-19 -{ m ~ INSERT 3A-13a - 3.1-19 ! The proposed cha e would dele the Action o place control rods in manual and, record RCC,,, hourly i f ultiple DRP per group are operable. fMultiple'in erable DRPIs of themselv , have no impac n SDM in MODES 1 and 1 2 if the ontrol rod p tions are v ified by altern e means (e.g., mo ble incore etectors). e requiremen to place contro rods in manual m not be appr priate in al situations a may be detrime al for load rejec 'on t nsients uni s operator ac on is assumed t simulate the rod ntrol system in a matic. Accid ts analyzed usi the [ Revised The al Design Procedure RTOP)] assume at the control ods are in [their ost limiting f mode). Automatic rod vement can acco odate a 10% load jection. Pi ing rd in manual may i act the load re ction capability ssumed when t P-9 tpoint was esta shed at 50% RTP The steam dump stem can acc modate a ] 40% RTP load re ction and with t rod control sys m in automat , a 50% RTP f load rejectio can be accommoda ed without a rea r trip. Whil manual operator a ion can be just timely as autom ic rod contr , there i no need to ve this limitati in the Technica Specificatic Correc ve acti for excessive ro motion are cover under ITS .7 Condit on C. The requirement to monitor and record T.,, h rly is unnec sary give he available indicator and alarms, e.g T.,, T,,, d e v a t i o n a l a , to alert operators to changing moderator conditions. ^ ~ ) l l i l l l 1

v ~ CONVERSION COMPARISON TABLE - CURRENT TS 3/4.1 P ne a no t TECH SPEC CHANGE APPLICABILITY NUMER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWlY 13 02 The requirement for inoperable digital rod position Yes Yes Yes Yes LS-15 indication is changed from "with a maximum of one per bank-to "one per group for one or more groups'. 13 03 A 4-hour Completion Time is specified to verify rod Yes Yes Yes Yes LS-12 position after movement of a rod with inoperable indicators more than 24 steps in one direction. 13 04 A requirement would be added to bring the plant to MODE 3 Yes Yes Yes Yes M within 6 hours if the required actions and completion times were not met. 13-05 The proposed change would retain an action statement. No. See CN 13 08-No. See CN 13 Yes Yes A currently in the plant TS. that permits continued POWER LS-20. LS-20. OPERATION with more than one digital rod position indicator 1 per group inoperable. 13 06 The change would allow separate condition entry for each No See CN 13 08-No. See CN 13 Yes Yes A inoperable DRPI per group or each demand indicator per LS-20. LS-20. bank. 13-07 The proposed modifications to the SR would verify agreement Yes Yes Yes Yes i M between digital and demand indicator systems prior to 1 criticality after esch removal of the reactor vessel head I instead of every 12 hours. The Frequency change is based on traveler TSTF-89. 13-08 Adds provision. in Callaway*s current specifications as Yes Yes No. Already in No. Already in current TS. current TS. LS-20 revised which, under certain conditions, would allow continued operation with more than one inanerable DRPI ner group. @ g epastste pnrtYh MerA 4d-gfJC $ -f CP-3.s-2Ol ggg i [3. re No njurtant W Jt5s 13-09 hrr~ent I .1.b$ b.lc~ fL dele . SDM ensur in a by siti WA MA 4 .3_ g, g 1ple i abl Is wi have impac on SOM S nd 2 i he cont r siti are ver ed b 7 p alt te s and r mot is lie const ent w g acci t anal ion of hese resent is k A/o / (4s ed.- j consi ent with aveler _W5 <mmrocalJ 5/1387 [ WCGS-Conversion Congparison TaNe-Cu 314.1 f

i \\ \\ l l N0 SIGNIFICANT HAZARDS CONSIDERATION (NSHC) CONTENTS I. O rg a ni z a t i o n........................................................... 2 II. Descri ption of NSHC Evaluations......................................... 3 III. Generic No Significant Hazards Considerations A Administrative Changes.............................................. 5 R Relocated Technical Speci fications.................................. 7 LG Less Restrictive (Moving Information Out of the Techni cal Speci fi ca ti ons)........................................ 10 M More Restrictive Requirements....................................... 12 IV. Specific No Significant Hazards Considerations LS LS 1.................................................................... 15 LS 2.................................................................... 17 LS 3.................................................................... 19 ) LS 4.................................................................... 21 LS.5........... ........................................................ 23 LS 6.................................................................... 25 LS.,........................................................... i LS.8......................................................... M..a._s.s..t... LS.9....................................................................35osa-nl M LS 10................................................................... 38 LS 11..............................................................Not Used LS 12................................................................... 40 LS 13.............................. .................................... 43 LS 14.............................. .................................... 45 LS 15................................................................... 47 LS.16........................................................ No t a ppl i c abl e LS 17........................................................ Not appl i c abl e LS 18................................................................... 49 LS 19........................................................ Not a ppl i c abl e LS20........................................3M......_C......".,.'. ppl i cabl e -....Not a LS 21.........................................( yy. - D Gs.1-23L LS 22............................................................ m. . 5_1 L_S_.23.._......._........_._.......... .. F..... M._....... A_ 5 9 -j 4 'J A - R l t Cu M w se.cr 4-b TR aa-cos i I V. Generic Technical NSHCs i TR.2.................................................................... 55 TR 3......................................... .......................... 57 I i WCGS-NSHCs-CTS 3M.] 1 S/l5/97 1 i l t

Detekc. A 3 I~l9 IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 23 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The proposed chan would delete the Actions to place control rods in ma 1 and record RCS T,, hour if multiple DRPIs per group are inoperable Actio b.1.b) and b.1.c) of LCO 3.1.3.2. Hultiple inoperable DRPIs will have no impac on SDH in Modes 1 and 2 if the c trol rod positions are verified by alterna means (e.g., movable incore detectors. The requirement to place control rod in manual is not appropriate in all situat ns and may be detrimental for load jection transients unless operator action is suned to simulate the rod contro system in automatic. Accidents analyzed using the Revised Thermal Design Proc re (RTDP)] assume that control rods are in [their no limiting mode]. Automat rod movement can acconinodate a 10% load rejectio. The requirement to nitor and record T,, hourly is unnecessary given the available dicators and alarm, e.g., T.,, Tm deviation alarm, to alert operators to chan ing moderator co itions. This proposed TS change has been ev uated and t has been determined that it involves no significant hazards consideration. This termination has been performed in accordance with the criteria set forth in CFR 50.92(c) as quoted below: "The Comission may make a fina etermination, pursuant to the procedures in 50.91, that a proposed amen tt an operating license for a facility licensed under 50.21(b) or 5.22 or or a testing facility involves no significant hazards consid ration, i operation of the facility in accordance with the proposed amen t would not: 1. Involve a signi icant increase in he probability or consequences of an accident prev ously evaluated; or 2. Create th possibility of a new or di ferent kind of accident frorn any acciden previously evaluated; or 3. Invo e a significant reduction in a mar 'n of safety." The following valuation is provided for the three catego ies of the significant hazards cons eration standards: 1. the change involve a significant increase in the obability or c sequences of an accident previously evaluated? Over 1 protection system performance will remain within the bou s of the previously pe ormed accident analyses since no hardware changes are propos WCGS-NSHCs-CTS 3M.] S3 S/1S/97

l 'D e te.ke. Q 3.\\- R ~ IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS s NSHC LS 23 (continued) j The reacti ty transients analyzed in USAR Section 15.4 will be una ected since rod position wil be ascertained to be consistent with those will no ffect the probability o any event initiators nor will the proposed cha affect the ability of any safety rela equipment to perfonn its intended functio. There will be no degradation in t perfonnance of nor an increase in the n r of challenges imposed on safety related 1pment assimed to function during a accident situation. Therefore, the propo change does not involve a sign icant increase in the probability or cons es of an accident previousi evaluated. l l 2. Does the change er te the possibility of new or different kind of accident l from any accident pr viously evaluated? There are no hardware changes r are ther any changes in the method by which any safety related plant system rfonas ts safety function. This change will not l affect the normal method of pla oper ion. No new accident scenarios, transient precursors. failure mec , or limiting single failures are introduced as a result of this cha Therefore, the proposed change does not create the possibility of a new or diff ent ind of accident from any previously evaluated. I 3. Does this change inv ve a signific t reduction in a margin of safety? The proposed change doe not affect the acce ance criteria for any analyzed event. There will be no eff on the manner in which afety limits or limiting safety system settings are determ nor will there be any e ect on those plant systems necessary to assure the acc lishment of protection functi s. There will be no impact on any margin of safet. NO SIGNIFICANT HAZARDS CONSIDERATI DETERMINATION Based on he above evalicion, it is concluded that the tivities associated with NSHC " 23" resulting 1;, the conversion to the improve TS format satisfy the no sign icant hazards consideration standards of 10 CFR 50.92y): and accordingly, a no si ificant hazards consideration finding is justified i l WCGS-NSHCs-CTS 3N.1 54 5/15/97

INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS i TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS Special Test Exceptions 1 3.1.10 is retained and renumbered as 3.1.8, consistent with this traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. f M 8 8 807 l TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. (TSTF-107MD Incorporated 3.1-6 [43.s-#r] TSTF-108[ N;t if;;.7,;ri:;d- --NA-- Net NRdapproved as of 1 Rev.1 incorporated 3.\\ - 2.1 t=ir ; ;;%d ;;. ra s.i oai l TSTF-110 Incorporated 3.1-10 +

• \\*#

Rev. 2. M *P9" TSTF-136 Incorporated 3.1-9,3.1-15 (b1Rc h M ITa3.s-coc.] TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoti date TSTF-142 N;Ui.wM..fe3 -NA-T.Teir Laeed e';cr# "I g;gd,;;MP42groh incorpora#4A

3. i - 22. _

Mn g / L 157 A (G 31-19 l (WOG21kN~ _ incorporated E 3.1-16 -MZ. ey "p /q s. - 2s J ~k l S/15/97 l

Rod Position Indication 3.1.87 Ms359d 3.1 REACTIVITY CONTROL SYSTEMS 3.1.87 Rod Position Indication LCO 3.1.8Z The Digital Rod Position Indication (DRPI) System and the $t8757t!$ Demand Position Indication System shall be OPERABLE. APPLICABILITY: H00ES 1 and 2. ACTIONS ..................................... NOTE Separate Condition entry is allowed for each inoperable rod position indicator gg@ and each demand position indicator gr krA. r, ,.y ED [q3.I-s9) CONDITION REQUIRED ACTION COMPLETION TIME A. One ptPI per group A.1 Verify the positicn of Once per Pl$f$$$$ inoperable for one or the rods with 8 hours more groups. inoperable position indicators Yhdirectly pg3112gg by using movable incore detectors. E A.2 Reduce THERHAL POWER 8 hours to s 50% RTP. WCGS-Mark-up ofNUREG.1431-ITS 3.1 3.1 17 S/15A7

Rod Position Indication 1.87 ggi 9g 3,t pidde co.&ol rab Erde imma;,b monaa as. no 3,2 M M & M ru.si945 kg. Once pv \\ hoar \\Q3.1-8l ACTIONS (continued) v - CONDITION REQUIRED ACTION COMPLETION TIME %<./3 VidFutMhitioli&f OriceX8 ggggg( B.:. MDNTikTnTorfCDRPDef hour.s ~~ BC9Jt4 ram @123e5fotione @ the2tels Ytt) oE*REOME 11ggermie:gattpn $3nu2% it!dt~cirtorL3ndirectly-brus'ing movable incotemegoo; 8HQ B.74 Restore;1ngge_rgle esitic!tQ)lodicaitors 2j4 i!!ours Q3.1-M] to?LNRWE31%)gus siTch"TiktMeWaiaMf onemaer2 group 1,s iR2perable; B d.One or more rods with B-C.1 Verify the position of 4 hours gggj inoperable p;siti;n the rods with MM@$ i indicator; @@If'have inoperable position been moved in excess of indicators indirectly l$3nM2d 24 steps in one by using movable direction since the last incore detectors. determination of the rod's position. QB B C.2 Reduce THERHAL POWER 8 hours to s 50% RTP. I WCGS-Mark-up ofNUREG-1431-ITS 3.1 3.1 18 S/2S/97

Rod Position Indication l B 3.1.87 i BASES l ACTIONS M (continued) simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. U Reduction of TliERMAL POWER to s 50% RTP pats the~ core into a condition where rod position is not significantly affecting core peaking factors (Ref. 32). The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above. Ed-@ [ o J.1-11 [scers3.1 MLee.~gi5rth~dhibi1RPEpefTitr.qigCftils22helpoti.tial 2ffZ4M ggKsTca E4,DB3!e]Isteisi_njiQyZujW5fitlBW + e e :, MW2GCSIE!IBK9!EagtfdenE4!oftaDIG8tt1N9EEBUR!balt!t Nap _veme$lf08 tangs.2FintDDr raMMR$3jali$ treeagamigerly;iniWeEthn3 4WretrB38tt'cMElEERB BilRBljagggggTfggtt$ cat lglOK.NE31DllIlli!. . c.i - il CLaLtgtitemmesfr.jourgioleg!stemullowt!!s!Kfam!d st1retistasetumDr'wy241tiourtwth.Js ggggnierof 10!ELig!nJ!! ants 2!anreignoreigntrictJLt1Eoxcryngienzaji!!]m i M!!'Dj!Ihewtb;t_haXMsitio(%ssM11? 0.1 ar.d 0.0E r bnd C:2 These Required Actions clarify that when one or more rods with inoperable psitior, f r. dis. tor; EIs have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 er,d A.0,:Juin!B712js WMi@le; are still appropriate but must be initiated promptly under Required Action B-1 CJ1 to begin inditectly verifying that these rods are still properly positioned, relative to their group positions. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.148 S/15/97

INSERT 8 3.1-48 0 3.1-19 Placing the Rod Control System in manual assures unplanned rod motion will not occur. The immediate Completion Time for placing the Rod Control System in ma'nual feflects the urgency with which unplanned rod motion must be prevented s while in this Condition. Monitoring and recording reactor coolant system T.,, help to assure that significant changes in power distribution and SOM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions, l 1 l l

1 I CHAME IMBER JUSTIFICATION l I statement, it may be possible for those unfamiliar with the DRPI J design to interpret the LCO as applying to all channels of DRPI. j 3.1 6 ITS LC0 3.1.4 would be split into two separate statements to clarify that the alignment limit is separate from OPERABILITY of the control rod. The COEITION A wording is broadened from "untrippable" to " inoperable" to ensure the CONDITION encompasses all causes of inoperability. Previous wordirn was ambiguous for j rods that, for instance, had slow drop times but were still trippable. These slow rods are inoperable rods, and the change clarifies the appropriate ACTIONS. The Bases are changed to reflect the changes to the LCO and COEITION A. These changes are based on traveler TSTF 107. 3.1 7 This change to the ISTS would incorporate, into LCO 3.1.7, an l"' " I i Action Statement that was previously approved _as part of __the j ~ e Callaway and Wolf Creek licensing basis.(p f'pastLAfrEpeMsure_ @ The Action Statement would permit continlied POWER UrtxATION t for up to 24 hours with more than one Digital Rod Position Indicator per rod group inoperable. The Action Statement specifies additional required actions beyond those applicable to the condition of one DRPI channel per group inoperable. The Bases for this change also would be incorporated into the Bases y $3l'MI for the plant._ITS.)fhe cha ~ ar onsist t with a er -4EXVDOef.). Jfie eu the IONS cha be s x consisMt witn t R ired stjnnt 3.1 8 The Frequency for ITS SR 3.1.7.1 for comparing DRPI and group demand position would be changed from 18 Months to "Once prior to criticality after each removal of the reactor vessel head." This change makes it clear that the surveillance must be performed each time the head is removed and that it is not tied to an absolute time interval. This change is based on traveler TSTF-89. ~ This change would eliminate ISTS 3.1.2 because the SDH 3.1 9 requirements for MODE 5 have been incorporated into Specification 3.1.1,in accordance with traveler TSTF 136. Traveler TSTF 9. Rev.1, relocated values for SDM to the COLR which removed the only difference between ISTS LCO 3.1.1 and ISTS LC0 3.1.2. Differences above and below 200*F will be addressed in the COLR. Subsequent sections have been renumbered. 3.1 10 Several surveillances (e.g., rod position deviation monitor and rod insertion limit monitor in this section) contain actions in the form of increased surveillance frequency to be performed in the event of inoperable alarms. These actions are moved from the WCGS-Differencesfrom NUREG-1431-ITS 3.1 2 S/lS87

l j ~ - l CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 rese 2.r3 . 'l SECTION 3.1 i DIFFERENCE FRON I G EG-1431' APPLICABILITY l It9BER DESCRIPTION DIABLO CANYON CONANCE PEAK WOLF CREEK CALLAINY t 3.1 7 An Action Statement that was previously approved as part of Yes Yes Yes Yes the current licensing basis of Callaws and_ Wolf Creek would be added to improved TS 3.1.7. a ~ 1 /G 3.1-39l [ T_joloprpy The Action Statement would permit operation [ Tor up Yo 24 hours with more than one Digital Rod Position l Indicator per group inoperable. j t 3.1 8 In accordance with traveler TSTF-89, the requirement to Yes Yes Yes Yes l compare DRPI against group demand position would be required whenever the reactor vessel head is removed, not l every 18 months. I j 3.1-9 This change would eliminate 1515 3.1.2 because the SDM Yes Yes Yes Yes requirements for MODE 5 have been incorporated into Specification 3.1.1 in accordance with traveler TSTF-136. 3.1 10 Several surveillances (e.g., rod position deviation monitor Yes Yes Yes Yes and rod insertion limit monitor in this section) contain actions in the form of increased surveillance frequency to be performed in the event of inoperable alarms. These l actions are relocated from the TS to licensee controll l documents. This is consistent with traveler TSTF-110. -{ M *11-004 } f 3.1 11 Not used. N/A N/A N/Al N/A l l 3.1-12 The Required Actions for inoperable DRPI are revised per Yes Yes Yes Yes i the current licensing basis to note that the use of movable [ f incore detectors for rod position verification is an indirect assessment at best. The position of some rods can j not be ascertained by this method. l l WCGS-Conversion Conperison TnNe-ITS3.1 S/15)97 l l 4 I j

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-3 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.2.1 Heat Flux Hot Channel Factor CTS 3/4.2.2 Heat Flux Hot Channel Factor (All FLOG Plants) DOC 02-06-A JFD 3.2-12 ITS SR 3.2.1.1 & 3.2.1.2 Frequency Comment: The ITS SR frequency has been changed from the STS frequency of 12 hours to 24 hours. This is based upon the incorrect justification that the CTS would allow 24 hours based upon ITS SR 3.0.3, since the CTS does not specify a frequency. Adopt the STS SR frequency of 12 hours. FLOG RESPONSE: (original) The change descriptions (DOC 2-06-A & JFD 3.2-12) will be revised to provide a basis for the 24 hours that is predicated on the time required to perform the surveillance. DOC 2-06-A is also revised to be DOC 2-06-M because this change is more restrictive than the CTS. Callaway and Wolf Creek are incorporating this change (DOC 02-06-A, JFD 3.2-12)in lieu of maintaining CTS which did not specify any completion time. DOC 02-13-LG (applicable to Callaway only) and JFD 3.2-17 are no longer used. FLOG RESPONSE: (supplement) As discussed in a telecon with the NRC staff on October 1, 1998, additionaljustification for the basis of the 24 hour surveillance frequency has been added to JFD 3.2-12. Additionally, this item is related to Comment Number O 3.2-7 for Callaway and Wolf Creek. No additional response is required for Comment Number O 3.2-7. ATTACHED PAGES: Attachment No. 8, CTS 3/4.2 - ITS 3.2 Encl.6A 3 l l l l

~ 'Ths, noto. W. Freepuwc3 -b sR 3.2.4.2. cou r2Vided ' conu~otent w% typicai pc.3entatiov > & m.ds ihat pov id < L-a p&,o d c4 -r A e. a f ck e.sts tre.h.Q con A s bon s - ~- CHANGE NUMBER ' JUSTIFICATION 3.2 12 bta o WCfE Seetonverefo'n Cem arEon +aNe) ca e ( os im.ERT g -s o q,, z,3, cas.s-,j 3.2 13 This 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 obtairst prior to r uiring that the surveillance be completed. {se.Ar c.,A-:5 -Q 3,2. q. 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 required to reduce power, establish equilibrium conditions, and obtain a flux map. 3.2 15 This change incorporates industry traveler TSTF-109. Action A.2 would require the QPTR be determined rather than performing a specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowgce that SR 3.2.4.2 may be performed in lieu __of SR 3.2._4.1.) -. fcr ". 2.2.'.? E&d te :gre grierreree if w.D m,. "^r -^re" ^^". ' g @ : tc jresicobicf These c1anges are 4412-io [ acceptable because they clarify the ISTS regarding frequency and use of incore flux monitoring for QPTR measurement. The changes reflect that incore detectors provide an acceptable QPTR determination during all plant conditions. 3.2 16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that is within its limit. F (Z) is approximated by a F a(Z) (which is obtained via SR 3.2.1.1) and F"a(Z) (which is obtained via SR 3.2.1.2). Thus, both F'a(Z) and F"a(Z) must be established to verify Fa(Z). This change is consistent with traveler WOG 105. 3.2 17 EFr ncy r remen or perfo ng F,measur nts h n g revi to e orm to S which not specify a omplet nT C ent tice i o perfo measdrement as s as ract The S SR Comp ionTjmesare sed o what sa no y rea able Comp ion Time for formin a f1 map: ever, i problems cur, the plant be f ed t reduc power o shutdown. his would subj the nt to tran ent cond on witho sufficie/dsafe basis There re, ma taining current TS requ ement s accep ble cause i @ useA. / Eqs.2.3 1 WCGS-Differencesfrom NUREG-1431 - ITS 3.2 3 S/258 7

INSERT 6A-3a 0 3.2-3/3.2-7 s The required time for completion of a flux map for determination of the heat flux hot

• channel factor is changed from 12 hours to 24 hours after achieving equilibrium conditions.

The proposed change affects SR 3.2.1.1 and SR 3.2.1.2. J A flux map is taken af ter a power level increase greater than a specified amount to verify F0 is within limits and to provide assurance that F0 will remain .within limits until the next required flux map is taken. Based on plant experience, the flux maps taken during power ascension provide a high degree of confidence that F0 will be within limits at the next power plateau. As such, the exact time period allowed for performance of the surveillance, after t reaching equilibrium, is not a significant safety consideration. The proposed time (24 hours) is a reasonable time period for obtaining and evaluating a flux map and then completing the procedural steps associated with this surveillance. Further, the 24 hour time period provides a reasonable limit on the length of time that the plant can operate in an unconfirmed condition. INSERT 6A 3b 0 3.2-4 The note was incorporated to address the rare situation where, during a mid-cycle shutdown, through further review of the previous surveillance, it was determined that the surveillance was invalid: or the required surveillance l frequency is not met due to the shutdown.. The amended Note would be required to return the reactor to a power level at which a new surveillance could be performed. l 1 I l i i i s ' l

ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: CA-3.5-002 APPLICABILITY: CA, CP, DC, WC REQUEST (original): Revise ITS 3.5.4 Bases to indicate that the RWST LCO, by virtue of its l temperature, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses). l REQUEST (revised): Revise various additional ITS Bases regarding the correct application of Criterion 2 of 10CFR50.36(c)(2)(ii). These changes are consistent with the attachment to a May 9,1988 letter from T.E. Murley (NRC) to R.A. Newton (WOG) entitled "NRC Staff Review of NSSS Vendor Owners Groups' Application of the Commission's interim Policy Statement Criteria to Standard Technical Specifications."

1. Revise ITS 3.5.1 Bases to indicate that the Accumulators LCO, by virtue of its pressure, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses).

2. Revise ITS 3 5.4 Bases to indicate that the RWST LCO, by virtue of its temperature, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses).

3. Revise ITS 3.6.7 Bases to indicate that the Recirculation Fluid pH Control (RFPC) System, by virtue of its TSP-C depth limit which ensures a minimum equilibrium sump pH of 7.1, also satisfies Criterion 2 (initial conditions of accident analyses). (Callaway only)

4. Revise ITS 3.7.6 Bases to indicate that the CST (and FWST for DCPP) LCO, by virtue of its water volume limit, also satisfies Criterion 2 (initial conditions of accident analyses).

ATTACHED PAGES: 1, CTS 3/4.5 -ITS 3.5 Encl. 5B B 3.5-4, B 3.5 29 3, CTS 3/4.7 - ITS 3.7 Encl. SB B 3.7-43 l l l l

= Accumulators B 3.5.1 BASES APPLICABLE For both the large and small break LOCA analyses, a nominal contained SAFETY ANALYSES accumulator water volume is used. The contained water volme is the (continued) same as the_available deliverable volume for the accumulators.. ;inc; th; occa;;;uletor; erc ca,.,ticd. Once di;ch;rged. Per mall t,rc;k;. en incrc;;c in water v;1=c i; a g;k clad tcagrature Fn;1ty. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volme from the accumulator to the check valve. The ;efcty ;ncly;i; ;;;ma; valuc; cf [5?S0] gallen; and [5070] gallen;. To allow for instrument inaccuracy, anla~ccumulator_voltme ranging betweer. valuc; cf 6520 6122 gallons and 0020 5594 gallons-ere is specified. The minimum boron concentration ;ctpint limit is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump bo.ron; concentration for post LOCA shutdown and an increase in the maximum sep pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. The large and small break LOCA analyses are performed at the minimm nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The effects on containment mass and energy releases from the acc ators are accounted for in the appropriate analyses (Refs. / 8 and /c4 3,s -co1] The accumulators satisfy Criterion 3 of the NRC Policy Stat = cat 10 CFRE50. ~.36..'~(cM2)(iih 24 yemj (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.S B 3.5 4 5/1S/97

l RWST B 3.5.4 l BASES s APPLICABLE The upper limit on boron concentration of 2200-25_00 ppm is withi.n the SAFETY ANALYSES val.ues used to determine the maximtsn allowable time to switch to hot (continued) leg recirculation following a LOCA._ The purpose of switching from cold leg to hot leg (lrjectica recirailatio@is to avoid boron precipitation in the core fo1 Towing thelaccident, L*ymydu\\>hid b MuMJr IO 3 G G *ll In the CCCS :n:1ysis, In the ainimum. containmentTpreissurelana]ysis for ECCS perfonnance, evaluation, the containnent spray temperature is e assumed to be equal to the RWST lower temperature limit of 35-37'F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the r;t: at which stcea :n be vcated cut the brc:k coreJoodifiDateTand increases peak clad temperature. The upper temperature limit of 100*F is used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment. The RWST satisfies Criterion 3 of the NRC Policy Stetaxat 10_CFR 50.363it)(2)(ii). LC0 The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical followir.g a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode. To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs. l APPLICABILITY In H00ES 1, 2, 3, and 4 RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since (continued) WCGS-Mark-up ofNUREG-H31 - Bases 3.5 B 3.5 29 S/1587

CST B 3.7.6 BASE.S i APPLICABLE contained water volume limit 1_ncludes an allowance for water.not SAFETY ANALYSES useablejbecause.of _ tank _ discharge line 1ocation orlother_ physical. (continued) racteristics. " Additional details regarding theidesign.~of_ thelAFW tem can'be~found in USAR 10.4.9. b x, 5.1 co M ~ The CST satisfies [ Career 6r3)and 4 of the imC NRC Policy Stataent 10 CFR 50.36'(c)(2)(11). c Ats i>--2,3} CA 3.5 -00 2-) LC0 To satisfy accidcat analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for four; hours following a reactor trip from 102% RTP, and then to cool down the RCS to RHR entry conditions, assuming a coincident loss of offsite power and the :nost adverse single failure. In doing this, it ;u:t retain sufficicit water to casur; adequatc act pc;itivc suctica hc;d for tra AIJ pump; during cocidown, ;; well as acccant for any lo:s;; frem tra stc;; drivan APJ pump turbinc, or beforc isciating APJ to a broken Hfte-The CST level required is equivalent to a usable volume of a 281,000 gallons, which is based on holding the unit in H00E 3 for 4 hours, followed by a cooldown to RHR entry conditions at 50*F/ hour. This basi.s is established in Reference 4 and exceeds the volume required by the accident analysis. The OPERABILITY of the CST is determined by maintaining the tank level at or above the minimum required level. APPLICABILITY In H00ES 1, 2, and 3, and in F00: 4, when stc;m gerarator is being relicd upon for Paat rcacval, the CST is required to be OPERABLE. In H00ES 4, 5, or 6, the CST is not required because the AFW System is not required. ACTIONS A.1 and A.2 If the CST level is not within limits, the OPERABILITY of the backup ESW supply should be verified by administrative means within 4 hours (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.7 B 3.7 43 S/15/97

.-= l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: NR 5.0-001 APPLICABILITY: CA, CP, DC, WC REQUEST: The NRC requested the following: For the following plants (and CTS sections), the applications identify the CTS requirements are being relocated to the FSAR: CW (6.2.3, ISEG; 6.5, review and audit; 6.10.1, record retention); CP (none); DC (6.10.1, record retention); and WC (6.2.3, ISEG; 6.5, review and audit; 6.8.2.3, procedure changes; 6.10.1, record retention). We discussed relocations to the QA plan with Ray Smith (QA branch) several weeks ago. The staff needs to have the licensees identify that these requirements are going to the QA plan and thus controlled by 50.54(a) The DOCS for relocating the above CTS sections are 1-04-LG and 3-09-LG. These DOCS only state the relocation is to the FSAR. The relocation should be to the QA plan. FLOG RESPONSE: Enclosure 3A and 3B has been updated to reflect the location of subject relocated items. ATTACHED PAGES: 1 ' 8, CTS 6.0 - ITS 5.0 1 Encl.3A 8 Encl. 3B 1, 7 l 1 I I i i l l

~.. CHANGE NUMER 82]C DESCRIPTION 1431. Rev. 1 to delete the term " Annual" and modify the submittal date. This change provides a reference to 10 s CFR 50.36a since 10 CFR specifies that the report must be submitted annually and include the results from the previous 12 months of operation. 03 07 A CTS [6.9.1.6], " Annual Radiological Environmental Operating Report" is revised to include specific details concerning the contents of the report. This change is consistent with NUREG 1431, Rev. 1. 03 08 A CTS Specifications [6.9.1.8, 6.9.1.9 and 6.9.2] are revised to delete the reference to submittal location for the monthly report, core operating limits report and special reports. The requirements related to report submittal are contained in 10 CFR. Since conformance to 10 CFR is a condition of the license, specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant requirements remain the same, the change is considered an adminstrative change. This change is consistent with NUREG 1431. Rev. 1. {#sbees\\ 03 09 LG The record retention reouirements are moved to MML apcMaphaent'100-prpoeduces] The removal of this detail from the CTS is consistent with NUREG 1431. The

a. hcese.

requirement for retention of records related to activities co.vtvolled doc.umed affecting quality is contained in 10 CFR 50, Appendix B. ~ Criteria XVII and other sections of 10 CFR 50 that are applicable to the plant (i.e., 50.71, etc.). Post-completion review of records does not directly assure operation of the facility in a safe manner, as the activities described in the documents have already been performed. By retaining these requirements (ta-pJanti Q -lig5MENFU]DifED licensee controlled document @Fany changes in these record retention requirements will be adequately controlled under the provisions of 10 CFR and the applicable regulations- ~ so. ii4 (a.) 03 10 LG The Radiation Protection Program is moved to the USAR consistent with NUREG 1431. This program requires procedures to be prepared for personnel radiation protection consistent with 10 CFR Part 20. These procedures are for the protection of nuclear plant personnel and have no impact on nuclear safety or the health and safety of the public. Requirements to have procedures to implement 10 CFR Part 20 are contained in 10 CFR 20.1101(b). Periodic review of these procedures is l l WCGS-Description of Changes to CTS 6.0 8 S/258 7 l

L 9 i CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 1 of 8 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK W011 CREEK CALLAWAY 01-01 The " Responsibility" section is revised to delete the Yes Yes Yes Yes A requirement to issue a management directive annually (i.e.. control room command function). The TS already adequately defines the function and, therefore. the management directive is redundant. C1 02 The " Plant / Unit Staff" section is revised to reflect the Yes No. CTS already Yes Yes A shift crew composition thble removal (if applicable), non-incorporates licensed personnel, and changes made to the section to be changes. [ on a unit basis vs. plant basis. Various editorial changes are made to accomplish the removal of the table and revisions to be consistent with NUREG-1431 and current plant practice. 01-03 The requirement for an SRO to be present during fuel Yes No. Deleted per CTS Yes Yes Amendment 50/36 i A handling and to supervise all core alterations is not retained in ITS. This requirement essentially duplicates f-[NR 5 O" Ooh the regulation in 10 CFR 50.54(m)(2)(iv). 01-04 The details of the review and audit. the independent safety No. Deleted per LAR No. Deleted per CTS Yes. Hove to USAR(f Yes. Move WJ5AR-LG engineering group, and training functions are removed from 117/115. Amendment 50/36 (ggDLa%,) A W E)#1oAeix m the CTS. Those items not specifically covered by a Chart'ar:1.f ESAR. regulation are moved to licensee controlled documents; j h, ML i, F otherwise the requirements are deleted. ( g 01 05 The requirement for the presence of an RO or an SR0 in the Yes Yes Yes Yes i [ A control room is delt.ted from the TS since the requirement is adequately controlled by 10 CFR 50.54(m)(2)(iii). 'i 01 06 The details regarding the minimum shift crew requirements Yes. Hove to FSAR No. CTS already Yes. Hove to USAR. Yes. Move to FSAR. l LG have been removed from the CTS because they are redundant contains changes. l to 10 CFR 50.54(k). (1), and (m) with the exception of the [ requirement for non-licensed operators. The minimum shift crew requirements will be moved to a licensee controlled document. SASM WCGS-Conversion Consparison Table - CTS.1M.0

s CONVERSION COMPARISON TABLE - CURRENT TS 6.0 ,Page 7 of 8 l ? TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALUWAY i I 03-08 CTS Specification [6.9.1.8. 6.9.1.9 and 6.9.2] are revised Yes Yes Yes Yes A to delete the reference to submittal location for tM [ reports. sec. p h lled ( monthly report core operating limit t_and special__ , doca _j K 6 0-8033 The_ record retention r luirements are moved tkbeeTM N-QA Ran M Yes -h4 Plan in Yes -QA Plan E Yes -QA PtJnh 03-09 'mMeantw!rhcoctdures] The requirement for retentton of l Cnapter rtof 5:EWL. P L'1*f h (Jugev rf f h Chap % r!of h. l LG i records related to activities affecting quality is g,, psAR, p [ ~ w i contained in 10 CFR 50. Appendix B. Criteria XVII and other t sections of 10 CFR 50 that are applicable to the plant (i.e. 50.71, etc.). 03 10 The Radiation Protection Program is moved to the USAR. Yes No. Deleted from Yes Yes i LG This program requires procedures to be prepared for CTS per Amendment l personnel radiation protection consistent with 10 CFR Part 50/36 l

20. Periodic review of these ;;;ccMares is required by 10 I

CFR 20.1101(c). h I 03-11 The High Radiation Area section is revised to be consistent Yes Yes Yes Yes A with the new Part 20 requirements. Changes are non-technical to add clarification. I l 03 12 The Process Control Program (PCP) section is proposed to be Yes. Move to FSAR. No. Deleted from Yes. Move to USAR. Yes. Move to FSAR. LG moved outside the CTS. The PCP implements the requirements CTS per Amendment [ of 10 CFR 20. 10 CFR 61, and 10 CFR 71. 50/36 ~ i 03-13 The following report [s] will be added to the ITS Yes Yes Yes Yes l M Administrative Controls section: " Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)* and [~ Post Accident Monitoring (PAM) Report".] 03-14 Shutdwn margin values would be moved to COLR per Yes No. Already part Yes Yes M traveler TSTF-9. In addition, moderator temperature of CTS. coefficient limits would also be moved to the COLR. [ i ( SASi97 l WCGS-Conversion Comparison Table - CTS 3M.0

=_. -.~. Att: chm:nt 2 to ET 98-0087 Prgs 1 of i LIST OF COMMITMENTS 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 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) 364-8831, extension 4077. I COMMITMENT Due Date/ Event A supplement to Reference 3 will be provided at a later date. Prior to issuance of SER. l l I l i I J l B l I N -}}