ML20154S237: Difference between revisions
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temperature, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses). ) | 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 ; | 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. Muriey (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." | Criterion 2 of 10CFR50.36(c)(2)(ii). These changes are consistent with the attachment to a {{letter dated|date=May 9, 1988|text=May 9,1988 letter}} from T.E. Muriey (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). | : 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). | : 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). | ||
Latest revision as of 00:08, 10 December 2021
| ML20154S237 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/21/1998 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20154S232 | List: |
| References | |
| NUDOCS 9810270355 | |
| Download: ML20154S237 (35) | |
Text
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i POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE,:nd ::p M: Of d: : " + ; th ;rtre /g-g/_Lp Tod ;;;iti;r,; i tM r, _ 1: ;tcp;. APPLICABILITY: MODES 1and27 O'O' A A ACTION: grouf
- a. With a maximum of one digital rod position indicator per%ek- /S-o.2-LS inoperablegiger,:
- 1. Determine the position of the nonindicating rod (s) indirectly Wi%fn 1
4- Aours "by.:di;t:1., the movable incore after any detectors motion of theatnonindicating least ence perrod 8 hours whichand/3-03-45 exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours or
- 3. te in Her r-rm,sspi4hin +4e nex+ 6 ha<r. /3-o+-M
- b. With more than one digital rod position indicator pergbe=+-inoperable I either: -
group (3-6-A l 1.a) Determine the position of the nonindicating rods indirectly by the movable incore detectors at least once per 8 hours and w/MTn f [ourJ iT:di;;;1., after any motion of the nonindicating rod which /S-03-/_5 l exceeds 24 steps in one direction since the last detennination l of the rod's posi . l 4.M -:'i;;;: :p:::^;a. ty,: rt{o:n,cdr 1 m bud : ntr;l cnd li-it rod cC:n 7,7 4 e / au.'Mer es g g,/ /9
' i nitor and e:r-d o :: :- S C :rt !;/:ta :.;;.re;; t: p:r:te: , , , , ,,
ic;n ca.c N hen . ;.nl < - - < w (T avg .. A _
/, b 1.d' flestore the digital rod position indicators to OPERABLE status I within 24 hours such that a maximum of one digital rod position I indicator perAbeek is inoperable, or /9-4S'A!
h4fd 2. BeinHOTSNDYYwithinthenext6 hours. With a maximum of one denand position indicator per bank inoperable l
- c. l either:
- 1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER
- 3. l3-0Y~)4 Y TYDY"5$)$ b8f b.)i-Hu*n fke nex+ b lur.c.
N' 1NIERT 2l+ /-I~] lg -s6--A CALLAWAY - UNIT 1 3/4 1-17 Amencment Nc. 61 _ Next pace is 3/4 1-176 9810270355 981021 PDR ADOCK 05000483 P PDR
l: 1 CHANGE NUMBER ElfC DESCRIPTION plant shutdown to Mode 3 requirement would be required in one less hour. 13 05 A This proposed change would involve retaining an action statement, currently in the plant TS, that permits continued POWER OPERATION with more than one digital rod position indicator per group inoperable. This is in accordance with the current licensing basis of the plant. 13 06 A Consistent with NUREG 1431 Rev. 1, a separate condition entry allowance is permitted in current TS 3.1.3.2 for each inoperable rod position indicator and each demand position indicator. This is an administrative change since the Required Actions address each rod or bank with inoperable indication separately [and Action b addresses the condition of multiple inoperable DRPIs, up to a complete loss of digital position indication]. 13 07 M The proposed modifications to the SR would require a verification of agreement between digital and demand indicator systems prior to criticality after each removal of the reactor vessel head, instead of every 12 hours. This reflects a reorganization of surveillance requirements in the ITS. The requirement for a 12 hour comparison would be moved to SR 3.1.4.1 in the ITS. The post vessel head removal requirement would be a new specification that demonstrates rod position system OPERABILITY based on a comparison of indicating systems. The Frequency requirement of prior to criticality after i each removal of the reactor vessel head would permit this ! comparison to be performed only during plant outages that involve plant evolutions (vessel head removal) that could affect the OPERABILITY of the rod position indication systems. The Frequency change is based on traveler TSTF 89. 13-08 LS 20 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 3B). 13 09 LS 23 - Cur ACTIONS b.1.b) and b.1.c) of LCO 3.1 deleted. 50 d in MODES 1 and ' rod position. Multiple inoperable DRPIs o impact on SDM in MODES 1 and 2 if the c rod ns are verified by alternate means rod motion is limited c ent with the acci analyses. Deletion of these requiremei \ co ent with traveler '40 73. Rev 1. - TJT/~ 234. l -,_ -- --,- Ns+ u cd. e z. H 9 DESCRIPTION OF CHANGES TO CURRENT TS 12 5/15/97
- - INSERT 3A-12 The proposed ch ge would delete the Actions to place contr ds in manual and 4 3./- /1 record RCS T,y ho ly if multiple DRPIs per group are ino ble. Multiple inoperable l DRPls, of themselves, ave no impact on SDM in MO 1 and 2 if the control rod positions are verified by ernate means (e.g., movabHfincore detectors). The requirement to place contr rods in manual may (be appropriate in all situations and may be detrimentalfor load r etion transient niess operator action is assumed to simulate the rod control system automati . Accidents analyzed using the[ Improved Thermal Design Procedure (ITDP) ss e that the control rods are in [ automatic).
Automatic rod movement can acco odate a 10% load rejection. Placing rods in manual may impact the load rejec ' n pability assumed when the P-9 setpoint was established at 50% RTP, The s am du system can accommodate a 40% RTP load rejection and with the rod co rol system i utomatic, a 50% RTP load rejection can be accommodated without a actor trip. While anual operator action can be just as timely as automatic rod contr , there is no need to h e this limitation in the Technical Specifications. Corr ive actions for excessive d motion are covered under ITS 3.1.7 Condition C. The equirement to monitor end reco T,y hourly is unnecessary given the available indi ors and alarms, e.g., T,y - Tn,e devia ' n alarm, to alert operators to changing m erator conditions. 4 Nrf tued. & 2.l-19
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.e CONVERSION COMPARISON TABLE - CURRENT TS 3/4.1 Page 8 of 10 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 13-05 The proposed change would retain an action statement. No. See CN No. See CN Yes Yes A currently in the plant TS, that permits continued POWER 13-08-LS-20. 13 08-LS-20.
OPERATION with more than one digital rod position indicator per group inoperable. 13-06 The change would allow separate condition entry for each No. See CN No. See CN Yes Yes A inoperable DRPI or each demand indicator. 13-08-LS-20. 13 08-LS-20. 13 07 The proposed modifications to the SR would verify agreement Yes Yes Yes Yes H between digital and demand indicator systems prior to criticality after each removal of the reactor vessel head instead of every 12 hours. The Frequency change is based on traveler TSTF-89. 13-08 Adds provision. from Callaway's current specifications as Yes Yes No. Already in No. Already in LS-20 revised which, under certain conditions, would allow current TS. current TS. continued operation with more than one inoperable DRPI per 9'0*P WWWW }I'fY~fW4F}( h S.l-20 W ,K{p4bG 13-09 ent TS ACTIONS b.1.b) and b.1.c) of LCO 3.1. -Nc r itin wiima Mc. ";;; in cu.im. -Ves-N[j] -Ves-LS 23 delet DM is ensured in HDDES 1 a rod position. Hultiple i le DRPIs wi e no inpact on SDH in -TS-- N[/} -TS- N[/} H0 DES 1 and 2 if t rol rod positions are verified by alternate mea rod mo limited consistent with I Muth, k 8 /~~! the ac analyses. Deletion o - requirements is stent with travelar ' ~ " -7'r-r - 4 M ,, _
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IV -s ee itnw=&e=rrs 14-01 Relocates current Specification 3.1.3.3 to licensee Y:.o'n; ti^ 55- j Yes. Relocated to No. See OL No. See OL R controlled document. -detetH0/4/95- g/" *TRM. Amendment No. 89. Amendnent No.103. OCL K -222),- l> C-/1LL-Do 4-15-01 S: ";d 0. @ Tim Smiikaiin 5.1.S.4 13 ieixaicd Y:: ' x: L "" ^S 07 -- Fa . Not " ;urrent Fe. Scc OL - Me. 5 : OL - -
- tdde of the Tednit:1 Sp^ctricati:n:. Thc RCS - d t-d 10/'/^S. TS. Svc CN
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CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97
.g a I ENCLOSURE 4 I N0 SIGNIFICANT HAZARDS CONSIDERATION (NSHC) CONTENTS I. Organization ........................................................... 2 II. Descripti on of NSHC Eval uations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 ) III. Generic No Significant' Hazards Considerations
'A Administrative Changes.............................................. 5 R Relocated Techni cal Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 LG Less Restrictive (Moving Information Out of the Techni cal Speci fi cati ons) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 M More Restri cti ve Requi rements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 IV. Specific No Significant Hazards Considerations LS LS 1...................... ............................................. 14 LS 2.................................................................... 16 LS 3..................................................................... 19 i LS 4................................................................ . 21 ,
V' LS-5............................................................. .... 23 ! LS 6.................................................................... 26 I LS 7............................................................. y ) *. 29 LS 8..................... ................................... ..
.-s 4 LS 9........................ ........................................... 32 LS 10................................................................... 35 LS 11..............................................................Not Used LS-12....................................................................37 LS 13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N o t Appl i c a bl e LS 14................................................................... 40 LS 15................................................................... 42 LS 16 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N o t a p pl i c a bl e LS 17 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N ot a p pl i ca bl e
- LS 18 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . No t a p pl i c a bl e LS 19 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N ot a ppl i c a bl e LS 2 0 . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N o t a ppl i c abl e LS 21......................................... ......................... 45 2..... .. ... .. ........ ....
LS 23............ v............... '............. ........... m....d......./..r.e. ..60- .. N ~N Ls-:4.- g2 y y,pu.3 V, Generic Technical NSHCs R-2............................................................... eft TR-3......................... ....... .... ..... .... ...... ........... & 54 1 5/15/97
FETE & 3,/-l9 IV. SPECIFIC N0 SlGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 23 1 10 CFR 50.92 EVALUATION ! FOR l TECHNICAL CHANGES THAT IMPOSE LESS RES CTIVE RE0VIREMENTS WIT 111N THE TECHNICAL SPEC ICATIONS The proposed hange would delete the Actions to pla control rods in manual and record RCS T., hourly if multiple DRPIs per group e inoperable. Actions b.1.b) and b.1.c) of LC0 3 1.3.2. Multiple inoperable DRPI will have no impact on SDH in Modes 1 and 2 if he control rod positions are erified by alternate means (e.g., movable incore de ctors). The requirement t place control rods in manual is not appropriate in all ituations and may be de imental for load rejection transients unless operator acti n is assumed to simul te the rod control system in automatic. Accidents analyzed us'ng the [ Improved T rmal Design Procedure (ITDP)] assume that
)
I control rods are in [a tomatic]. Aut tic rod movement can accommodate a 10% load rejection. The requir nt to monito and record T., hourly is unnecessary given the I available indicators and larms, e.g , T,, Tm deviation alarm, to alert operators to changing moderator con tions. This proposed TS change has n evaluated and it has been determined that it i involves no significant hazar consideration. This determination has been i performed in accordance with criteria set forth in 10 CFR 50.92(c) as quoted i below: l "The Corrmission ma make a 'nal determination pursuant to the procedures in 50.91. that a pro osed amen nt to an operating license for a facility licensed under 5 .21(b) or 50. or for a testing facility involves no significaat ha ros considerati . if operation of the facility in accordance with the prop ed amendnent woul ot:
- 1. Invol e a significant increas in the probabi1ity or consequences of an acc ' dent previously evaluated; r
- 2. C eate the possibility of a new o different kind of accident frorn any ccident previously evaluated; or
- 3. Involve a significant reduction in a gin of safety. "
The foll ing evaluation is provided for the three cate ries of the significant hazards consideration standards:
- 1. Does the change involve a significant increase the probability or consequences of an accident previously evaluated?
Overall protection system performance will remain within th bounds of the previously performed accident analyses since no hardware chan es are proposed. The reactivity transients analyzed in FSAR Section 15.4 will b unaffected N0 SIGNIFICANT HAZARDS CONSIDERATION 50 5/15/97
b eun~e &2Hg IV. SPECIFICN0SIGNIFICANTHAZARDSCONSIDERATIONS[ NSHC LS 23 (continued) sinc rod position will be ascertained to be consist t with those analyses. The pr sed change will not affect the probabilit of any event initiators nor will the proposed change affect the ability any safety related equipment o perform its intended function. T re will be no degradation in the perfo nce of nor an increase in the n r of challenges imposed on safety relat equipment assumed to functi during an accident situation. Therefore, the roposed change does not i volve a significant increase in the probability or nsequences of an accid t previously evaluated.
- 2. Does the ch ge create the po bility of a new or different kind of accident from ny accident eviously evaluated?
There are no hardware anges r are there any changes in the method by which any safety-related plan syst performs its safety function. This change will not affect the norma hod of plant operation. No new accident scenarios, transient precu ors, failure mechanisms, or limiting single failures are introduced a a esult of this change. Therefore, the proposed change does not create pos ibility of a new or different kind of accident from any previously ev uated.
- 3. Does this cha ge involve a gnificant reduction in a margin of safety?
The proposed cha does not affect t e acceptance criteria for any analyzed event. There w 1 be no effect on the anner in which safety limits or limiting safet system settings are dete ined nor will there be any effect on those plant s stems necessary to assure t accomplishment of protection functions. here will be no impact on any rgin of safety. N0 SIGNIFICANT HAZARDS CONSIDERATI DETERMINATION Based on the ve evaluation, it is concluded that the etivities associated with NSHC "LS 23" resulting from the conversion to the improve TS format satisfy the no significan hazards consideration standards of 10 CFR 50.9 c): and accordingly, a no signif cant hazards consideration finding is justified.
/
NO SIGNIFICANT HAZARDS CONSIDERATION 51 5/15/97
Industry Travelers Applicable to Section 3.1 - ' ;. TRAVELER # STATVS DIFFERENCE # COMMENTS TSTF 9, Incorporated 3.1 1 NRC approved. Revision 1 l TSTF 12. Incorporated 3.1 15 NRC approved. ITS Revision 1 Special Test Exception 3.1.10 is retained and l re numbered as 3.1.8. consistent with this traveler and TSTF 136. TSTF 13. Incorporated 3.1-4 NRC approved. Revision 1 TSTF 14 Incorporated 3.1-13 NRC approved. Revision 4.4- 7%-7. /-445-TSTF 15, Incorporated NA NRC approved. Revision 1 TSTF 89 Incorporated 3.1 8 NRC approved. TSTF 107/,y.) Incorporated 3.1 6 f .T. /-/.5 TSTF 108, -Net-)fhcorporated -NA- -Net NRC approved..as-ef-Revision 1 7. /-2 / +-mm,_ em+ m<< ++m 7 -7 /w/ TSTF 110 Incorporated 3.1 10 A/gc ofjoro v,/, Revision-1-2 N'I /~#8Y TSTF 136 Incorporate'd 3.1 9, 3.1 15 W--7. /-Bol, A//C affrJV8d TSTF 141 Not incorporated NA Disagree with change; traveler issued after cut off date. , TSTF-142 -Net-/licorporated -NA- ,NK
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WOG 105 Incorporated 3.1 16 - MARK-UP 0F WOG STS REV 1 (NUREG 1431) 5/15/97
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Rod Position Indication iki-B 3 R 3.1 9
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3.1 REACTIVITY CONTROL SYSTEMS 3-heEE7 Rod Position Indication LCO 3-1-6 M The DStabG Rod Position Indication (gRPI) System and the B PS Demand Position Indication System shall be OPERABLE. APPLICABILITY: H0 DES 1 and 2. ACTIONS l
........................................N0TE - - - -
Separate Condition entry is allowed for each inoperable rod position indicator per 2.1 7 . group and each demand position indicator. g r L A. g CONDITION REQUIRED ACTION COMPLETION TIME A. One llRPI per group A.1 Verify the position of the Once per 8 hours B PS. inoperable for one or rods with inoperable 3.1 12 more groups, position indicators m by using movable incore detectors. DE A.2 Reduce THERHAL POWER to 8 hours s 50% RTP. (continued) MARKUP OF WOG STS REV 1 (NUREG 1431) 3.1 17 5/15/97
. a Rod Position Indication i 3-1-0 E12 3.1 9 l o ACTIONS (continued) 1 CONDITION REQUIRED ACTION COMPLETION TIME XNS6A'T 3.1-l? & 3/-/f E kg!$tgggggglPJg -g 4 : . .
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u.. . c .. i.1 .. EhC2 One or more rods B-g1 Verify the position of { hours B with inoperable the rods with DRIGN pe;ition inoperable position 3.1 17 indicators have indicators M 3.1 12 been moved in by using movable incore excess of 24 steps detectors. in one direction since the last E determination of the rod's position. Ehg2 Reduce THERMAL POWER to 8 hours , s 50% RTP. I (continued) l MARKUP OF WOG STS REV 1 (NUREG 1431) 3.1 18 5/15/97
- . . - . . . . .- -. - -.-.. .-.. ........- .- _.. - - ..~.-.-. .
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- i. INSERT 3.1-18 Q 3.1 19 i 1
l REQUIRED ACTION COMPLETION TIME 1 B.1 . Place the control rods under manual control. Immediately ) AND B.2 Monitor and record RCS T y. Once per 1 hour 1 AND l
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. ; .s BASES ACTIONS L1 (continued) l l simultaneously having a rod significuntly out of position and an
- event sensitive to that rod position is small.
d LZ l Reduction of THERMAL POWER to s 50% RTP puts the core into a i condition where rod position is not significantly affecting core l peaking factors (Ref. 3). i The allowed Completion Time of 8 hours is reasonable, based on i operating experience, for reducing power to s 50% RTP from full j- power conditions without challenging plant systems and allowing ! for rod position determination by Required Action A.1 above. l l mujudulet;8.7, e~/ A+ 4 7.H1 Acry'/lebY
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(continued) MARK UP OF NUREG 1431 BASES B 3.1 48 5/15/97
1 , INSERT B 3.1-48 Q 3.1-19 a The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. I Monitoring and recording reactor coolant system Toy help to assure that significant changes in power distribution and SDM 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 l i i i
CHANGE NUMBER JUSTIFICATION rods are inoperable rods, and the change clarifies the appropriate ACTIONS. The Bases are changed to reflect the changes to the LC0 and CONDITION A. These changes are based on traveler TSTF 107. v 3.1-7 This change to the ISTS would incorporate. into ITS LC0 3.1.7. an l Action Statement that was previously approved as part of the Callaway and Wolf Creek licensing basis.;d6C 'y:57pThe Actiondf. /-/1 1 Statement would permit continued POWER OPERATION for up to 24 hours l with more than one Digital Rod Position Indicator per rod group inoperable. The Action Statement specifies additional required actions beyond those applicable to the condition of one DRPI per group inoperable. The Bases for this change also would be incorporated into , I the Bases for the plant ITS. -The cham ar^ c--"'--t
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3.1-8 The Frequency for ITS SR 3.1.7.1 for comparing DRPI and group demand I position would be changed from 18 Months to "Once prior to criticality after each removal of the reactor vessel head." This change makes it clear that the surveillance must be performed each time the head is removed and that it is not tied to an absolute time interval. This change is based on traveler TSTF 89. 3.1 9 This change would eliminate ISTS 3.1.2 because the SDM requirements for MODE 5 have been incorporated into Specification 3.1.1 in accordance with traveler TSTF 136. Traveler TSTF 9. L.1. relocated values for7R'-3./-ogg SDM to the COLR which removed the only difference between ISTS LC0 3.1.1 and ISTS LCO 3.1.2. Differences above and below 200*F will be addressed in the COLR. Subsequent sections have been re numbered. 3.1 10 Several 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 TS to licensee controlled documents since the alarms do not themselves directly relate , to the limits. This detail is not required to be in the TS to provide ' l' adequate protection of the public health and safety. Therefore, moving ) this detail is acceptable and is consistent with traveler TSTF 11 , TA'-:P. /-oo.p. l 0 2: 2. 3.1 11 Not used. 3.1 12 The Required Actions for inoperable DRPI in ITS 3.1.7 are revised per the current licensing basis to note that the use of movable incore detectors for rod position verification is an indirect assessment at best. The position of some rods can not be ascertained by this method. 3.1-13 This change adds an 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 LC0 requirement and SR were added to verify that JUSTIFICATION FOR DIFFERENCES TS 2 5/15/97
CONVERSION COMPARIS0N TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 3.1 Page 1 of 3 TECH SPEC CHANGE APPLICABILITY NUEER CESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.1-1 In accordance with industry traveler TSTF-9fev-t- this Yes Yes Yes Yes change would relocate the specified limits for SDM from 77 .7./-Oo(o several TS to the COLR. 3.1-2 ~ s .... ";t: to SR 3.1.^.1. .h;d i ch .i? - u f 5 _ L /y', ~ Me-- N/jg e no n a o , o m , L H ;nio;niny
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10TC ;rdir.;. n; ,,,,,, ...,. y p yp m. n:rr!!::ti^ Of pr:dt;t;d i at;.itj =!re: t -^- raa-4 -- A- -, i -
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-te ;;; nf;d =!re: iM!' M de .; pri:r te _; ding : '= L ' '[
vu, nuy yc,;c = pur -u ~r*m. yd-ur,j. A 2- 4 3.1-3 -"ci! NC ~3 .1.0 Re gi d Acti n C.1 i: revi d- -Ne-/(/p -No-yp Wes-ji/p -No-/\/ gyg y Tivm % ii. M^^C vu 3. iv 'De ir, "00C 2 .iith (,, , 1.c." - 3.1-4 In accordance with industry traveler TSTF-13Av-t- ISTS Yes Yes Yes Yes 77-7,/-tD(, SR 3.1.4.2 which requires verifying MTC within the 300 ppm boron limit, is deleted and the Note in that SR is moved to the SR that requires the lower MTC limit to be verified. The deleted SR is not a requirement separate from the lower MTC verification SR but is essentially a clarification of when the SR for the lower MTC limit should be performed. 3.1-5 Per current TS [3.1.3.1]. the words "with all' are removed Yes Yes Yes Yes from the LCO for control rod alignment limits. This ensures that the number of channels of DRPI required to be OPERABLE will not be misconstrued. 3.1-6 In accordance with traveler TSTF-107, the change provides Yes Yes Yes Yes additional clarification that the alignment limits in the LCO apcarate frne tqERABILITY of a control rod. n 3.1-7 /An Action Statement that was previously approved as part of Yes Yes Yes Yes the current licensing basis of Callaway and Wolf Creek would be added to improved TS 3.1.7,7g/gy' $ T./-/9 J % 4 .2.<'The Action Statement would permit operation for up to 24 hours with more than one Digital Rod Position Indicator per group inoperable. _ j
}
~
% -- ~
% / _ x .
l CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMAT80N NO: CA-3.1-005 APPLICABILITY: CA REQUEST: Revise ITS SR 3.1.3.2 Bases to correct a typographical error. The Note 1 SR limit is 300 ppm, not 200 ppm. ATTACHED PAGES: Attachment 7, CTS 3/4.1 -ITS 3.1 Enclosure 58, page B 3.1-20 i
HTC B s-1-+ ggg BASES SURVEILLANCE SR 3.1.4.2 crd S 3.1.4.3 (continued) REQUIREMENTS N CA -3.1-60s
~ - 3 _ / "
g 3 If the 300 ppm Surveillance limit is exceeded, it is O S./,C-/ possible that the E0C limit on MTC could be reached before the planned E0C. Because the MTC changes slowly with core depistion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the E0C limit. 4g The Surveillance limit for RTP boron concentration of d 7 I 6'I 60 ppm is conservative. If the measured NTC at 60 ppm is F M g . orc positive than the 60 ppm Surveillance liait, the E0C limit will not be exceeded because of the gradual manner in which MTC changes with core burnup. REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.
- 2. FSAR, Chapter (MlIImh O #'I-
- 3. WCAP s e A . 92-73 "l' A, " Westinghouse Reload Safety Evaluation Methodology," July 1985.
- 4. I"I't. Chapter 25.
MARX UP OF NUREG 1431 BASES B 3.1 20 5/15/97
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. (supplement) DOC 2-06-A is also revised to be DOC 2-06-M because this change is more restrictive than the CTS. (original) 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 JDF 3.2-17 are no longer used. (supplement) As discussed in a telecon with the NRC staff on October 1,1998, additional justification 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 Q 3.2-7 for Callaway and Wolf Creek. No additional response is required for Comment Number Q 3.2-7 ATTACHED PAGES: Att. No. 8 CTS 3/4.2 - ITS 3.2 Encl. 6A 2, insert 3.2-12
CHANGE NUM5f.3 JJSTIFICATIM setpoint adjustments. and channel restoration), adding 2 hours for necessary initial preparations (procedure preps, calibration ecuipment checks obtaining tools and approvals), it is reasonable to expect a tctal of 18 hours. Further, setpoint changes should only be required for extended operation in this condition. Finally, the Bases for making this setpuint change is exactly the same as the NUREG Bases provided for the 72 hour Completion Time of LCO 3.2.1 Required Action A.4 which is also a setroint reduction. In summary this change is acceptable because it would permit time to perform required flux mapping, permit orderly resetting of the high flux trip setpoints, and reduce the chances of an inadvertent reactor trip during the required power reduction. 3.2 07 Consistent with TSTF 97, the NOTE in SR 3.2.1.2 is revised by removing the phrase "is within limits and" to clarify that the actions to be t taken if F (Z)e is increasing are required regardless of whether F a(Z) is within its limits. 3.2 08 Consistent with TSTF 99, the LC0 3.2.1 (F aMethodology). Required Action 3.1 Comoletion Time for t.".e recuction of ths AFD limits if F'a(O is not within limits is increased from 2 hours to 4 hours. This makes it consistent with the Completion Time associated with Required Action A.2 of LCO 3.2.1 (Fn Methodology). The change is acceptable because it eliminates an inconsistency in the ISTS. 3.2 09 For consistency with current TS 3.2.4 and improved TS 3.3.1 condition D the breakpoints for the applicability of the surveillances in the notes in improved TS SR 3.2.4.1 and SR 3.2.4.2 are modified to be applicable at less than or equal to 75 RTP and greater than 75t RTP. respecti velyrwi4-s-ac=mi-swat 4-vs r.hancA tha: retains current TS. requirements C d - Y]j '
%x M v. W M ~"sTF-1.11 3.2-10 Consistent with TSir 1 . this change moves requirements for increased surveillance frequencies in the event of inoperable alarms to licensee controlled occuments. This change is acceptable because it removes ;
requirements regarding alarms and alarm responses that are not necessary to be in the TS to protect public health and safety. 3.2 11 Not applicable to Ce w ay. See Conversion Comparison Table (Enclosure 65).
~
D LI 3 .1 ~ G_ -- 3.2 13 This change retains In'eWmance of peaking factor determinations following plant shutdowns. The CTS. througn the exemption to specification 4.0.4. allows prerecuisite plant conditions to be oc;ngr to recu1 ring that the surveillance be complete
.I s .3. 2 - / 3 j {"~
j J JUSTIFICATION FOR DIFFERENCES TS 2 5/15/97
Q 3.2-3/Q 3.2-7 INSERT 3.2-12 3.2-12 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. A flux map is taken after a power level increase greater than a specified amount to verify FQ is within limits and to provide assurance that FQ 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 FQ will be within limits at the next power plateau. As such, the exact time period allowed for performance of the surveillance, after 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. 1 l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.13-6 APPLICABILITY: CA REQUEST: ITS 3.4.13 Bases LCO a. (Callaway) Comment: The intent of the addition that leakage past instrumentation lines not being pressure boundary leakage is une. lear. Is that leakage upstream of isolation valves? If ! it is, is there a line size limit and is this consistent with the description of pressure boundary in the FSAR and the definition in ITS Section 1.17 I FLOG RESPONSE (original): This Bases change refers to 3/8 inch tubing for instrument connections to ASME Class 1 fluid piping downstream of the root valves and to 1/8 inch core exit thermocouple sheaths. These instrument lines are not part of the reactor coolant pressure boundary (RCPB), as discussed in FSAR Table 3.2-1 Notes (9) and (10). As further stated in Sub-article NCA-1130(c), the scope of ASME Section Ill does not apply to instrument tubing and that tubing is not designed or specified to be part of the RCPB or provide a pressure retaining barrier. As discussed in FSAR Sections 9.3.4.2.3.5 and 15.6.5.2, normal charging can accommodate a 3/8 inch break and maintain normal pressurizer level such that the ECCS is not actuated. This Bases change does not refer to leakage upstream of instrument root valves. There is no conflict with ITS Section 1.1. FLOG RESPONSE (supplement): As discussed with NRC on October 8,1998, the Bases has ; been revised to provide additional discussion. l ATTACHED PAGES: 0, CTS 3/4.4 - ITS 3.4 8, page B 3.4-84 l
j RCS Operational LEAKAGE j B 3.4.13 BASES (continued) j LCO RCS operational LEAKAGE shall be limited to: i Pressure Boundary LEAKAGE
- a. '
i No pressure boundary LEAKAGE is allowed, being indicative , of material deterioration. LEAKAGE of this type is j unacceptable as the leak itself could cause further i deterioration, resulting in higher LEAKAGE. Violation of l this LCO could result in continued degradation of the l 1 ' RCPB @
& 3.+.Gm-l l LEAKAGE past seals l end gasketsm .
l ' 1s not pressure boundary LEAKAGE.A {
. b. Unidentified LEAKAGE 1Auner 8 2+-r+ Q24/H l
\
One gallon per minute (gpe) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the ! containment air monitoring and containment sump level I monitoring equipment can detect within a reasonable time period. Violation of this LC0 could result in continued degradation of the RCPB. if the LEAKAGE is from the pressure boundary, c Identified LEAKAGE Up to 10 gpa of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of gidentified LEAKAGE and is well within the capability'of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). - - _ g $ M ./.?-3 ym--- ..
~Q G---E 3 wa_4 4.W- - ,E f
~~~ m ==~r"~~~S
' Violation of this LCO could result j in continued degradation of a component or system. -
(continued) l l MARX UP OF NUREG 1431 BASES B 3.4 84 5/15/97 i i
---.sa.- .an--....~n,.-~..~...~.. , . . - . .~-.~~-n .,
INSERT B 3.4-84 Q 3.4.13-6 Instrumentation lines are 3/8 inch tubing for instrument connections to ASME Class 1 fluid piping downstream of the root valves and 1/8 inch core exit thermocouple sheaths. These instrument lines'are not part of the reactor coolant pressure boundary (RCPB) nor do they provide a pressure retaining barrier. Normal charging can accommodate a 3/8 inch break and maintain normal pressurizer level such that the ECCS is not actuated. h
l l ADDITIONAL INFORMATION COVER SHEET 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 ! 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. Muriey (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 satisf5s 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: l Attachment 11, CTS 3/4.5 -ITS 3.5 ! Enclosure SB, pages B 3.5-5 and B 3.5-30 ! l Attachment 12, CTS 3/4.6 - ITS 3.6 Enclosure 58, page B 3.6-53 Attachment 13, CTS 3/4.7 - ITS 3.7 Enclosure 58, page B 3.7-44
Accumulators B 3.5.1 1 3* ( C ri-{e r io n APPLICABLE The accumulators sati y'triterion 3 of the "P4 Pc,licy Statscat CA45-W SAFETY ANALYSES plCFRE5_0363M(2)Jjip. _ -
^
(continued) LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA. the ECCS acceptance criteria of 10 CFR 50.46 (Ref.p could be violated. c) eg_,7, gag j For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 2000 [0_00 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist. This LC0 is only applicable at Bg pressures > 1000 psig. At pressures s 1000 psig, the rate of RCS blowdown is such that the i ECCS pumps can provide adequate injection to ensure that peak . clad temperature remains below the 10 CFR 50.46 (Rei. p) limit of 2200*F. [, anal 7 L3 cA_y,ggj In MODE 3, with RCS pressure s 1000 psig, and in MODES 4, 5, and 6. the accumulator motor operat .d isolation valves are closed Wlth]poWurkr.emoVe~d F:fr.5impheTvaTTeToWrltWrl to isolate the accumulators from the RCS IRefs R ; p.M . This allows RCS CA-7.5-$/ cooldown and depressurization without disc arging the accumulators into the RCS or requiring de ressurization of the accumulators. g c ,,m ,f,4 , 7 ,,f,.g g 7 , onlyreg u reed wh en die a e e u,,,u la n jo eenaee it re<4re on or aal de -He YNYNr$c S* *
$"/"[*jeI fenyer,+ur# e, ascontinued) al/swe y d2 MARK-UP OF NUREG 1431 BASES B 3.5 5 5/15/97
.. #At /* li+I+ cue wrj eevidd o
in e J 7t K.
, =
1 RWST , B 3.5.4
. n.
, NGES i
APPLICABLE assumes that all control rods are out of the core. M SAFETY ANALYSES m_ Op.F a , -n.&.s ,n awr: m:, (continued) w M_Di2 1 2'_ F - -
^-a u .2 A _ c; " . . . . . <
L e.r i >. -. u &%&.V ." s a :: .~. - y " . ..s
,(,).'.- f2
.l . g . ., ' , , ._. t .- , , - .,,r. , . . . . g -. s : - g 1- fg
. . .o . . w . ,,,,..;
...a:.
., . , . . ; ;g
;.'.s .\ y The upper limit on boron concentration of 2299 M ppa is used to determine the maxista allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg inj;;tica WNWWIMti is to avoid boron precipitation in the core following the accident, cwlwirmr# lukpwrepr44N *f +4e In theVECCS analysis, the containment spray temperature is CA-7.5-23 asstamed to be equal to the RWST lower temperature limit of 35 E' (. If the lower temperature limit is violated, the h*g*y)- con;ainment spray further reduces containment pressure, which j decreases the rate at which steam can be vented out the break and !
increases peak clad temperature. The upper temperature limit of M 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. LitigMigggMe p M. in v. 2 - G :.12 0 .c: W e t i e - Co-id-erim 3 and The RWST satisfiesVCriterion 3 of th; = F;', icy St;t;.st g _.7, gn LO.R.ER"JB56%!QggEE. __ x LCO 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 following a DBA, and to ensure adequate level in the containment (continued) MARK UP OF NUREG 1431 BASES B 3.5 30 5/15/97
. =
l Spr;y Additiv; Syst;; 'At ; phcric. Subet.T.;;pheric. I;; Ceadenser. nd Ouel) BeckculationiElkid[p[ Control 2 System B 3.6.7 f b BASES BACKGROUND 'ha Centeir.-.,; Sprey Sy;t;; ntu; tier, ;ignel ;p;ns tra velv;; l (continued) fr th; ; prey ;dditin tak to the sprey pu;p n;tiens er the ' cent;ir.;~r,t : prey pu;p -t;rt ;ignel sp;as the velas fr the ____.._222m.._ i__,. _,1__ _ e _2_.._ 2_,.. vu_ eo. dys uJ "w'88* m_ ,. u_m, 6" ' W5 6*' " * '"65 8 "J * 8-W 6V V 8'UV8 ' 21;tica i-m__. _ 2 2, m . . _ drar,'inte th; ; prey p5 n;tiens. The ; prey _____m.. ___.2;__ ,__ mu_ _222 _, u_m, __,..m__ , uwMIb5Yb bw.In bufub , bJ fl VW4Mbd I V5 b5Ub MWM .b5V,5 m__ V5 31MV3 5 dV 5 W b 3 V 3 _,, _s A ._ ... . AL_
.~ . . .
._A__ _____.._2 r..,
,___ A
. ,, - .L_
.~ ,
ne sey
, . . 2._..A._
___t_ b...2.._.-_.A
. . . v,t_
,~
l 1 l gr;;..; xiutica ;,4 volu;; ef zlutien ; prey-d int; sat;ir.~r.t I ca;urs ; len; t;r; ;;at;ic..~r,t :;u p p" of . 0.0 c.ad s 0.5. Thi cr,';ures the satinued i; dire r;tcatien effGtinais ef the I x;p w;ter during the recirculetia phe;; of ; prey egr; tion erd
;1,; ;ini;i ;; tre xarr;,,c; cf chierid; indund ;; tress arresier, c;; king of tre steinia stal rairsi; tion pipirg.
,,,j rehnh%
APPLICABLE The BEG Sprey Additive System is essential to the removal f SAFETY ANALYSES airborne iodine within containment following a DBA. C A - J.In-do 2 Following the assumed release of radioactive materials into ' containment, the containment is assumed to leak at its design value volume following the accident. The analysis assumes that MHHF 15HlWlq@@3 of containment is covered by the spray (Ref. 1). The DBA response time assumed for thegI g Sprey Additive System is the same as for the Containment Spray System and is discussed in the Bases for LCO 3.6.6 " Containment Spray and Cooling Systems." Tho' DBA analyses assume that one train of the Containment Spray System /Spr y Additiv; Sy;t;; is inoperable. and thet th; entir;
;p ey edditiv; t;nk v;l;;; i; ;;dded t; th; r; .ining Cent;ir. Tent av, ., as.._m.__ ,,_.__m.u..
. . - y. .
The RFE.C Svrey Additin System satisfies Criterie,A3 of the "PsCCA-35-dQ, , P;1 icy Stet;.;;at 101CE50{361c)t2)If1D Crsbr. 2 ed
/ ~ -
}
LC0 The RgG Spray Additive System is necessary to reduce the release of radioactive material to the environment in the event v a DBA. To be considered OPERABLE, the volume end ancentretica of ISR. ~~C - j the sprey odditiv; solutica must be sufficient t; provid; ";0;.
, (continued)
MARK.UP OF NUREG 1431 BASES B 3.6 53 5/15/97
- -4.- , ,
CST . B 3.7.6 BASES f- v N h APPLICABLE uC - J A W AG 6W w-M u .w x.UMHG .6 SAFETY ANALYSES (continued) gh.;;.-ic ' d ., CA-3.6-ad 2 The CST satisfies 3, of tk = = P;1 icy l Stetement har P.wyc wwe++ . I ( ^ ~ LC0 To satisfy exiir,t analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for IEEERmtB following a reactor trip from 102% RTP, and then to cool down the RCS to RH1 entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In d;ing , this, it ;;st r;tein sufficient ater t; n;ur; eiqu;t; ret I p;;itiv; ;uction had f;r tk 'J',,' pgs during ;;;1deu ,, 2 wil 4
. .. ..... ......,...m.,
..,,~.........m. yy ......m. I
;r Mf;r;ixl:tia; = t:;tr;unlix.
The . m H CST 4evel va m . A ' - a .. r; quired is equivel;nt to ; us;bi; velu.e of a 281,000 gallons, which is based on h;1 ding tk unit ir, = 3 fer i heur;, fell;;;d by a cooldown to RHR entry conditions . .
- s- ..-e. We%m et 7/heur. This basis is established in Reference 3 4-end en;;i th; volu ; required by tk x;ider.; ;nalysis.
The OPERABILITY of the CST is determined by maintaining the tank 4evet y W1GeWe '.a.C at or above the minimum required MM APPLICABILITY In H00ES 1, 2, and 3, and in = 4. uhcr, st; n gn;reter is being r li;d up;n for heet reavel, the CST is required to be OPERABLE. In H00EMI or 6 the CST is not required because the AFW System is not required. ACTIONS A.1 and A.2 If the CST 4evet m is not within limits, the OPERABILITY of the backup gg supply should be verified by administrative means within 4 hours and once every 12 hours thereafter, OPERABILITY of the backup f;;ieter K supply must (continued) MARK UP OF NUREG 1431 BASES B 3,7 44 5/15/97 1
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA-3.5-004 APPLICABILITY: CA REQUEST: Clarify ITS SR 3.5.2.3 Bases to reflect OL Amendment No.127 dated August 17, 1998 regarding ECCS pump venting. ATTACHED PAGES: Attachment 11, CTS 3/4.5 -ITS 3.5 Enclosure SB, page B 3.5-20
e - ECCS Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.2 REQUIREMENTS 1 (continued) Verifying the correct alignment for manual, power operated, and ' automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience. SR 3.5.2.3 Ecc.T ANA "d # ' __m _, , _u__ e , u. i t.. m.m.u m
.m _ or ..m. 2. _,_ ... a...__,........m..,_
_ _ . r..
. _ y_ .
2..__.. . . m.u . EC puhors are normally in a standby nonoperating mode. As @266-l l such, ow path piping has the potential o develop voids and pocke of entrained gases. Maintaining he piping from the ECCS s to tre RCS full of water h_ _-_ _ -- r r==n-g CA_ygg4 2=r_-_-_: c1115BleM ensures that the system will perform properly, injecting its full capacity into the RCS upon demand.j ggesls 5]E3110tQgglglgeg@pM"3HteterTtI8tcA 3f.g liefGbeNTwe:mo,4widusiWM8dtBtlRorplutit*Bei pggbh45.3,1 This w 1 also prevent water hasser, pump cavitation, and pumping of no ondensible gas (e.g., air, nitrogen, or , l hydrogen) into t reactor vessel following an SI signal or i during shutdown ling. The 31 day Frequency takes into consideration t gradual nature of gas accumulation in the ECCS piping and the rocedural controls governing system operation.
.ZNSEA*T 8 T.S@0 SR 3.5. .4 % -
Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other (continued) MARK UP OF NUREG 1431 BASES B 3.5-20 5/15/97
INSERT FOR B 3.5-20 CA-3.5-00f-1 The design of the centrifugal charging pump is such that significant noncondensible gases do not collect in the pump. Therefore, it is unnecessary to require periodic pump casing ! venting to ensure the centrifugal charging pumps will remain ' OPERABLE. f r\ 6--g k a
l 4 ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: NR 5.0-001 APPLICABILITY: CA, CP, DC, WC
, REQUEST: The NRC requested the following:
i 4 . 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. l The staff needs to have the licensees identify that these requirements are going to the l i 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: Enclosures 3A and 3B have been updated to reflect the location of the subject relocated items. ATTACHED PAGES: Attachment 18, CTS 6.0 - ITS 5.0 Enclosure 3A, page 8 Enclosure 3B, pages 1 and 6 i a
CHANGE NUMBER EC DESCRIPTION 03 06 A CTS [6.9.1.7], " Annual Radioactive Effluent Release Report" and CTS [6.14.c] is revised consistent with NUREG 1431. Rev. 1, to delete the term " Annual" and modify the submittal date. This change provides a reference to 10 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 Specification [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 administrative change. chan C s consistent with NUREG 1431. Rev. 1. This @R 5.0-o 03 09 LG The record retention requirements are moved to . .: F". ^a
- NU i;; incd;C 9 The removal of this detail
- h. h wha. c.hoM from the CTS is cons stent with NUREG 1431. The p Au,m .nt, requirement for retention of records related to activities 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 rformed. in these Wrt: record retention any M licensee controlled documen1 requirement c controlled under the provisions of 10 CFR @ and the applicable regulations. k5o.54 MD 03 10 LG The Radiation Protection Program is moved to the FSAR 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 ~ DESCRIPTION OF CHANGES TO CURRENT TS 8 5/15/97
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V Page 1 of 7 CONVERSION COMPARIS0N TABLE - CURRENT TS 6.0 APPLICABILITY TECH SPEC CHANGE C0HANCHE PEAK WOLF CREEK CALLAWAY DIABLO CANYON NUMBER DESCRIPTION Yes Yes Yes The " Responsibility" section is revised to be consistent Yes 01-01 A with current plant practice. The requirement to issue a , management directive annually (i.e.. control room comand function) is deleted. The TS already adequately defines the function. and therefore the management directive is redundant. No - CTS already Yes Yes Yes 01-02 The " Plant / Unit Staff" section is revised consistent with incorporates A current plant practice. Sections are revised to reflect changes the shift crew composition table removal (if applicable). i non-licensed personnel. and changes made to the section to be on a unit basis vs. plant basis. Various editorial , changes are made to accomplish the removal of the table and l revisions to be consistent with current plant practice. l No - Deleted per Yes Yes ; The requirement for an SRO to be present during fuel Yes A 01-03 handling and to supervise all core alterations is not CTS Amendment 50/36 ]J ! I retained in ITS. This requirement is deleted. This l requirement essentially duplicates the regulation in 10 CFR N 50.54(m)(2)(iv). [~ I No - Deleted per No - Deleted per Yes. move to USAR Yes moved to CAR 01-04 The details of the review and audit, the independent safety i-and-fMWt GA Mn CTS Amendment 50/36 r [ue LG engineering group and training functions are moved from the CTS. Those items not specifically covered by a regulation CTS Amendment 117/115 (s%fah//.,, an,/ - in chapkr G *(
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,~q CA pfsc /'7 4M FJAR. Reuw are moved to licensee controlled documents; otherwise the je;%
requirements are deleted. et #e //fAA. ,'^f Yes Yes j Yes Yes 01-05 The requirement for the presence of an RO or an SRO in the ~ - , A control room is deleted from the TS since the requirement is adequately controlled by 10 CFR 50.54(m)(2)(lii). No - CTS already Yes move to USAR Yes, move to FSAR The details regarding the minimum shift crew requirements Yes - Hove to FSAR 01-06 contains changes LG have been removed from the CTS because they are redundant 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. C #1 C 707 . . . . _ . . . . . . . . . - . . . . . . - . - t
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CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 6 of 7 , TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY , i 03-06 CTS [6.9.1.7]. " Annual Radioactive Effluent Release Report' Yes Yes Yes Yes A and CTS [6.14.c] is revised consistent with NUREG-1431 Rev. 1. to delete the term " Annual" and modify the submittal date. s 03 07 CTS [6.9.1.6]. " Annual Radiological Environmental Operating Yes Yes Yes Yes A Report" is revised to include specific details concerning the contents of the report. , 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 the -
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monthly report. core operating limits report and special -- ~ g reports. -- ~ go.IUm in. c M v M do avaenh 03-09 Therecordretentionrequirementsaremovedtoh: F7." c5D Yes- $/l //.n fn Yes-MA [/en In Yes- $/l [/.n in Yes- Q A Pttn LG C y - tt g p.u d e D The requirement for retention of riscords related to activities affecting quality is
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ffej g f:~rj)g 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.). 3 , 03-10 The Radiation Protection Program is moved to the FSAR. Yes No - Deleted from Yes Yes [NR 5.o-coq LG This program requires procedures to be prepared for CTS p Amendment personnel radiation protection consistent with 10 CFR 50/36 ; Part 20. Periodic review of these procedures is required by 10 CFR 20.1101(c). , 03-11 The High Radiation Area is revised to be consistent with Yes Yes Yes Yes A the new Part 20 requirements. Changes are non-technical to add clarification. 03-12 The Process Control Program (PCF) section is proposed to be Yes - move to FSAR No - Deleted from Yes - moved to USAR Yes - moved to the LG moved outside the CIS. The PCP implements the requirements CTS per Amendment FSAR , of 10 CFR 20.10 CFR 61, and 10 CFR 71. 50/36 03-13 The following report [s] will be added te to the ITS Admin Yes Yes Yes Yes H Controls Sectinn:
- Reactor Coolant System (RCS) Pressure and Terrperature Limits Report (PTLR)" [and ~ Post Accident ,
Honitoring (PAH) Report."] i CONVFQRTON COMPADTRON TARIr . OllDQFNT T9 G/1G/Q7}}