ML20153H043
| ML20153H043 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/24/1998 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20153H029 | List: |
| References | |
| NUDOCS 9809300293 | |
| Download: ML20153H043 (300) | |
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Attachm:nt 1 to ET 98-0078 , P ge 1 of 5 l l l l l JLS CONVERSION TO IMFROVED TECHNICAL SPECIFICATIONS l i CTS 3.4 - REACTOR COOLANT SYSTEM i ITS 3.4 - REACTOR COOLANT SYSTEM l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES l l l l i i 9809300293 DR 980924 p ADOCK 05000482 PDR ,_
Attachm:nt 1 to ET 98-0078 Prge 2 of 5 INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER 3.4. Gen-1 CA, CP, DC, WC YES 3.4.1-1 CA, CP, DC, WC YES 3.4.1-2 CA, CP, DC, WC YES 3.4.1-3 DC NA 3.4.2-1 CA, CP, DC, WC YES 3.4.3-1 CA, CP, DC, WC YES 3.4.4-1 CA, CP, DC, WC YES 3.4.5-1 CA, WC YF' 3.4.5-2 CA, WC Y 3.4.5-3 CA, WC YE , 3.4.5-4 CP NA 3.4.6-1 CA, CP, DC, WC YES 3.4.6-2 DC NA 3.4.7-1 WC YES 3.4.7-2 WC YES 3.4.7-3 CA NA 3.4.8-1 CA, CP, DC, WC YES 3.4.8-2 DC NA 3.4.9-1 CA, CP, DC, WC YES 3.4.9-2 CA NA 3.4.9-3 CP, DC, WC YES 34.9-4 DC NA 3.4.10-1 CA, CP, DC, WC YES 3.4.11-1 CA, CP, DC, WC YES 3.4.11-2 CA, CP, DC, WC YES 3.4.11-3 CA, CP, DC, WC YES 3.4.11-4 CA, CP, DC, WC YES 3.4.11-5 WC YES 3.4.11-6 CA, CP, WC YES 3.4.12-1 CA, CP, DC, WC YES 3.4.12-2 CA, CP, DC, WC YES 3.4.12-3 CA, CP, DC, WC YES 3.4.12-4 CA, CP, WC YES 3.4.12-5 CA, WC YES 3.4.12-6 DC NA 3.4.12-7 CP NA 3.4.12-8 CP NA 3.4.13-1 DC, WC YES 3.4.13-2 CA, DC, WC, CP YES 3.4.13-3 CA, CP, WC YES 3.4.13-4 CP,DC NA 3.4.13-5 DC NA 3.4.13-6 CA NA
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Att: chm:nt 1 to ET 98-0078 Pige 3 of 5 INDEX OF ADDITIONALINFORMATION (cont.) ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER 3.4.14-1 CA, CP, WC YES 3.4.14-2 CP, DC, WC YES 3.4.14-3 CA, CP, DC, WC YES 3.4.14-4 CA, WC YES l 3.4.14-5 DC,WC YES 3.4.15-1 CA, DC, WC YES 3.4.15-2 CA, WC YES 3.4.15-3 WC YES 3.4.15-4 CP,DC NA 3.4.15-5 DC NA j 3.4.16-1 CA, CP, DC, WC YES 3.4.16-2 WC YES 3.4.16-3 WC YES 3.4.G-1 CP YES CA 3.4-002 CA NA CA 3.4-003 CA NA CA 3.4-004 CA, CP, DC, WC YES CP 3.4-004 CP NA DC 3.4-ED DC NA DC ALL-001 (3.4 changes only) DC NA DC ALL-002 (3.4 changes only) DC NA DC ALL-005 (3.4 changes only) DC NA DC 3.4-003 DC NA TR 3.4-004 CA, CP, DC, WC YES TR 3.4-005 CA, CP, DC, WC YES TR 3.4-006 CA, CP, DC, WC YES TR 3.4-009 CA, CP, DC, WC YES WC 3.4-001 WC, CP YES WC 3.4-002 CA, CP, DC, WC YES WC 3.4-004 WC YES WC 3.4-006 WC YES WC 3.4-007 WC, CA, DC YES l WC 3.4-008 WC, CA YES WC 3.4-009 WC YES WC 3.4-0010 WC, CA YES I
-. .-- - - . . - - . . - .. ~ ._- - - _ _ . - - . ._ Att chm:nt 1 to ET 98-0078 P gs 4 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is f ollowed for submitting additional information:
- 1. Each licensee is submitting a separate response for each section.
- 2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.
- 3. If a licensee initiated change does not apply, "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.
- 4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,38, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as " PLANT SPECIFIC DISCUSSION."
- 5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.
- 6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.
- 7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the original license amendment request to reflect the changes being made by one or more of the other licensees. These changes are not included in the additional information for the licensee to which the change does not apply, as the changes are only for consistency, do not technically affect the request for that licensee, and are being provided in the additionalinformation being provided by the licensees for which the change is applicable. The complete set of changes for the license arnendment request will be provided in a licensing amendment request supplement to be provided later.
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Attichm:nt 1 to ET 98-0078 Pcge 5 of 5 j JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued)
- 8. The item numbers are formatted as follows:
(Source][lTS Section]-[nnn] Source = Q - NRC Question I CA - AmerenUE DC-PG&E WC - WCNOC CP - TU Electric ; TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL" is used for the section number. j nnn = a three digit sequential number l l
ADDITIONAL INFORMATlON COVER SHEET ADDITIONAL INEORMATION NO: Q 3.4. Gen-1 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 3.4.x Bases General l There have been a number of instances that the specific changes to the STS Bases are not properly identified with redline or strikeoat marks. Comment: Perform an audit of all STS Bees markups and identify instances where additions and/or deletions of Bases were nm properly identified in the original submittal. i FLOG RESPONSE: The submitted ITS Bases markups for Section 3.4 have been compared 1 to the STS Bases. Some differences that were identified were in accordance with the markup ] methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorial in nature and would not have affected the review. Examples of editorial changes are:
- 1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced. !
- 2) Changing a verb from singular to plural by adding an "s" without " redlining" the "s." l
- 3) Deleting instead of striking-out the A, B, C, etc., foliowing a specification title (e.g.,
SR3.6.6A.7).
- 4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference. i
- 5) In some instances, the brackets were retained (and struck-out) but the unchanged text l within the brackets was not redlined.
- 6) Not redlining a title of a bracketed section. The methodology calls for the section title to ,
be redlined when an entire section was bracketed. i
- 7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was determined to not be applicable, the text was then struck-out and remains in the ITS Bases mark-up.
Differences of the above editorial nature will not be provided as attachments to this response. The pages requiring changes that are more than editorial and are not consistent with the markup methodology are attached. ATTACHED PAGES: Encl. 5B B 3.4-15, B 3.4-17, B 3.4-19, B 3.4-21, B 3.4-22, B 3.4-25, B 3.4-27, B 3.4-28, B 3.4-34, B 3.4-38, B 3.4-39, B 3.4-42, B 3.4-47, B 3.4-48, B 3.4-49, B 3.4-52, B 3.4-55, B 3.4-64, B 3.4-68, B 3.4-69, B 3.4-71, B 3.4-72, B 3.4-74, B 3.4-75, B 3.4-80, B 3.4-87, B 3.4-88, B 3.4-98
RCS P/T Limits B 3.4.3 BASES ACTION B.1 and B.2 (continued) increased stress or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more i careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. In reduced pressure and temperature I conditions, the possibility of prepagation with undetected flaws is , decreased. !
! If the required restoration activity cannot be eccomplished within 30 minutes, Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.
If the required evaluation for continued operation cannot be accomplished within 72 hours or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1 and Required Action B.2. A favorable ' evaluation must be completed and documented before returning to operating pressure and temperature conditions. Pressure and temperature are reduced by bringing the plant to H00E 3 within 6 hours and to MODE 5 with RCS pressure < 0 psig within
^
36 hours. '
"QL K s.4.sen-M " * "'"
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without chall enging plant systems. C.1 and C.2 Actions must be initiated imediately to correct operation outside of
- the P/T limits at times other than when in MODE 1, 2, 3, or 4 so that the RCPB is returned to a condition that has been verified by stress analysis.
1 The imediate Completion Time reflects the urgeno of initiating action
- to restore the parameters to within the analyzeo i n,ge. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre analyzed transients in the stress analyses, or inspection of the components. ASME Code, Section XI, Appendix E (Ref. 7), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-15 S/1/5/97
1 l RCS Loops-MODE 5. Loops Filled B 3.4.7 t i ! BASES (continued) l l APPLICABLE In H0DE 5 RCS circulation is considered in the SAFETY ANALYSES determination of the time available for mitigation of the l accidental boron dilution event. Thc R"R iccp; provid; this
;irculatian.
The operation of one RCP in_H0 DES 3, 4,;and;5 provides adeguate flow to.. ensure mixing,lprev.ent~ stratification, and produce l gradual reactivity _ changes during 'RCS' boron concentration reductions.ENith(nofre. actor coolant]oop;inloperation in e.ither H0 DES ~3,4,~ _orj5,{ boron _ dilutions;must beiterminatedland' dilution sources isolat_ed. . 'MboronTdflutionEahalysiLinitheLHODES3ake credi.t; for the ' mixing volume assocfhtedjwith;having at least?one reactor ~ coolant loop in operation' '(Ref.1). l . RCS Loops-H0DE 5 (Loops Filled) have tc;n idcatified in the NRC Policy St;tc; cat as important centributors to risk reduction satisf1_es. criterion _4Lof 10_CFR 50.36(c)(2)(ii). LCO we The purpose of this LCO is to require that a least one of the
)
qu5 2 ' RHR loops be OPERABLE and in operation with additional 43,4so ., loop OPERABLE or two SGs with secondary side ater level 2 . E - One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under these conditions. An additional RHR loop is required to be OPERABLE to meet single failure considerations. However, if the standby RHR l
# N loop is not OPERABLE. an acceptable alternate method is two SGs with their secondary side range water levels a I 4,.u.s - 2.
~- ~ Should the operating RHR l@oop fail, the SGs could
. I pM ""' 58Tgg used to qu.s.3 b remove the decay heat via natural circulation. '
l thstrumentation rem 3 e 66 % MN NE
- W/ Note 1 permits all RHR pumps to be removed from operation for p7.5GWide raylml The purpose of the Note were -
uruv4@%k..)isdc cacrgiced s 1 hour per 8 hour period. to permit tests that are_ required to be perfomed without flow or pump noise. desigacd to validate various accidcat onely;;; valucs. Orc of the tests g rferred during th; startup testing progr;; is th; validation of red drop tire; during c;1d conditions, both with and without ficw. The no flow test may be perfor cd in ."00C 2, 4, or and rcquircs that thc pu7ps be stopped for ; ;hcrt griod ;f ti;c. The Not; g mits dc crcrgicing of the pumps in ordci t; grfor; this tc;t end validetc th; ;;;ced anelysi; valucs. If changc; cic ;;dc to the l RCS that would cau;; ; chang; to the fisw chorectcristic; cf the l RCS, thc input valuc; must bc rcialidstcd by conducting thc tc;t t egem- The 1 hour time period is adequate to perform the j necessary testing, and operating experience has shown that boron (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 6 3.4 33 S/1/5/97
i RCS Loops-MODE 5. Loops Filled B 3.4.7 BASES APPLICABILITY In MODE this LCO requires 5 forced withcircula RCSp.loops d fjeetIfilled N M (tion- ortne reac remove deny heat from the core and to provide proper boron i l mixing. One loop of RHR provides sufficient circulation for l these purposes. However, one additional RHR loop is required to i be OPERABLE, or the secondary side (6Ei5iibrange water level of at l least two SGs is required to be a i 667, Fd 3 34,g.2. ; 4 3.4.C .1 ' Operation in other MODES is covered by: - l LC0 3.4.4, "RCS Loops-H0 DES 1 and 2": LCO 3.4.5, "RCS Loops-H0DE 3": LCO 3.4.6, "RCS Loops-H0DE 4": LCO 3.4.8, "RCS Loops-HODE 5. Loops Not Filled"; LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (HODE 6); and LCO 3.9.6. " Residual Heat Removal (RHR) and Coolant j Circulation-Low Water Level" (HODE 6). l l i l l i l ACTIONS A.1 and A.2 4 A4 S'2' ( l 66*/. p 5 4.'s 3 l A If one RIR loop is rable and the required SGs have secondary l Wa r96) side
- water levels 4 redundancy for heat removal is lost.
l Action must be initiated imediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The imediate Completion Time reflects the importance of maintaining the i 1 availability of two paths for heat removal. B.1 and B.2 If no RHR loop is in operation, except during conditions i permitted by Notes 1 and_4, or if no loop is OPERABLE, all j operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. Addition ~o~f borated water with a concentration greater than or equal to the~ minimum required RWST concentration but less than the. actual RCS. boron concentration'shall not be considered a reduction in boron concentration. (Ref.:2). To prevent inadvertent; criticality duringLa boron dilution, forced circulation from:st3 east one RCP l 1s required to provide proper mixing. ;rd pic;cric the ;rgir,to critic;1ity ir, thi; ts; cf cpcretier.. The imediate Completion Times reflect the importance of maintaining operation for heat removal.
- (continued)
WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 35 5/U5/97 l
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I RCS Losps-H00E 5. Loops Filled B 3.4.7 l-L BASES (continued) l SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours that the required loop is in operation. Verification may include flow rate, i temperature, or pump status monitoring, which help ensure that i forced flow is providing heat removal. The Frequency of 12 hours l l 1s sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance. , i SR 3.4.7.2 uivale@ g I Verifying that at least two SGs are OPERABLE byMsuringtheirpu.sa( secondary side narrow range water levels are a Q@ ensures an alternate decay heat removal method is available via natural ' circulation in the event that the second RHR loofis ~not I L OPERABLE.a If both RHR loops are OPERABLE, this Surveillance is not needec . The 12 hour Frequency is considered adequate in view of other indications available in the control room to alert the operator t the loss of SG 1evel.
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i
~
The nm rangr level m humenWihn m*Oe. u *td- , A >taDes if h ; i SR 3.4.7.3 Ip scr were. eum one. tede. eq<. level- ih%icator-Verification that a second RHR pump is OPERABLE ensures gg additional pump can be placed in operation, if needed, to - that an %3 maintain decay heat removal and reactor coolant circulation. l Verification is performed by verifying proper breaker ali nt We. and oower available to RHR pump. If secondary side range water level is a in at least two SGs, this Survel Sance is not needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been L shown to be acceptable by operating experience. l l REFERENCES Nene- 1. USAR, Section 15.4.6 1
- 2. NRC letter (W. Reckley to N. Carns) dated November 22, 1993: " Wolf Creek Generating Station '-' Positive Reactivity Addition: ~ Technical Specification Bases l Change."
! 3. MtC Information. Notice 95 35, " Degraded Ability of SGs to . { Remove Decay Heat by Natural Circulation." j - .. - - i i l WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 36 S/1/5/97
CHANGE !MBER JUSTIFICATION plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim. there is no need to perform this SR within the 24 hour period specified in NUREG-1431 Rev.1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition. the THERMAL POWER specified in the Note would be changed from the generic value in brackets (90 % RTP) to 95 % RTP, This change is acceptable because it specifies a power level in better agreement with current operating procedures for performing a precision heat balance. Current TS do not specify a power level for this measurement. 3.4 41 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). 3.4 42 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). 3.4 43 A new Condition is added to LCO 3.4.1 to reflect the current licensing basis of Wolf Creek for RCS flow rate. License Amendment 61 approved revisions to incorporate the provisions of the RCS flow TS entitled "RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR" into the "DNB PARAMETERS" specification. These changes were made to support the use of VANTAGE SH fuel with the Intermediate Flow Hixer grid feature. This amendment also approved operation at an increased power level. I33A.G 2. x 4 c-2 3.4 44 t o WCMee f.omr6rsiop-Oc(pap +s6n Tatrie] cl re . leERT sA-8b y 3.4 45 i ITS 3. 2 has can re sed move he Not for equ ed Acti B.1 r gardin CP p ip swap operat ns dt ' licabi ty Not for a umulat isol ion th LCO. as dis ssed i ravel WOG5).Rev ecif 4 tim llowan s for xceeding'the LC sn [. .er P1f nt- [E S]
' ps cap
. e of jectingM nto tt)e'RCS pre i corpo ate .
as disc sed in N3.438). TheseNotpdde il s ons wher except nstoteLC0areperm}tted,e)ndare)ua e j a repriat y annotAfed undpr the LCO. NSE AT GA -8 a.-)- 3.4 46 Consistent with current TS 3/4.1.1.4, "Hinimum Temperature for Criticality." ITS LC0 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current TS wording is acceptable since valid II' CGS-Differencesfrom NUREG-1431 - ITS 3.4 8 S/lS/97
INSERT 6A-8a 0 3.4.12-2 Consistent with traveler TSTF-285, ITS 3.4.12 has been revised to move the Note for Required Action 8.1 regarding CCP pump swap operations and the Note under Applicability regardinc accumulator isolation to the LCO. These Notes have beer reworded for clarity and detail situations where exceptions to the LCO'are permitted. Also, piant-specific time allowances for exceeding the LC0 number of [ECCS] pumps capaDie of injecting into the RCS are. incorporated [, as discussed in CN 3.4-18). INSERT 6A-8b 0 3.4.5-2 0 3.4.5-3 Steam generator levels for MODE 3, 4, and 5 are specified to ensure SG tubes are covered. The current TS did not ensure tube coverage. l l l t I l l' l l
ezge 6 or a CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-143I SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION r.-- ptpr G ,;,td Mkibc.dCTS SR 4.2.3. 6,the. detMis on 44 enY* bed. which. h hjlagv ate 6 Mkd s's veW4_ Tvo_m ST& SM.4.t . A+m+. _ 3.4-38 tonsis wityisTF-1 . tha detai ,on t by Yes J* h k@ X {Q3 .4.1-1 j hwhi the RJJr' flow r e ar erif ar ao f SR3f.1.4to Ba .f L Yes Yes I The shutdown requirements of ITS 3.4.11 would require Yes Yes 3.4-39 ! the plant reduce T , to <500T within 12 hours, rather than H00E 4. to address the concern of entering [LTOP] LCO 3.4.12 Applicability with inoperable PORVs. For ! consistency. the shutdown requirements of ITS 3.4.16 are also revised to all 12 hours to reduce T , to
<500T. This change is consistent with TSTF-113.
No - See CN YM Yes The Note to SR 3.4.1.4 would be modified to specify a No - See CN 3.4-40 3.4-51 3.4-34 plant specific reactor power and to provide additional time to perform an RCS precision P rate measurement. No No No 3.4-41 LCO 3.4.1 is revised to reference Tables 3.4.1-1 and Yes - Allowance , added per 3.4.1-2 for RCS total flow rate limits for DCPP Units 1 and 2 respectively. Amendment 60/59. No No An exception to SR 3.4.14.1 frequency to leak test Yes - Specific No . 3.4 42 PIVs 8802A, 8802B and 8703 has been added. This to DCPP change is consistent with the DCPP current TS. No Yes No No 3.4-43 n new Condition is added to LCO 3.4.1 to reflect the current TS of Wolf Creek for RCS Flow Rate. No .50- Yes @a45-1 No 3.4-44 Steam generator levels for MODES 3. 4 and 5 are g $ N 'S _i specified to ensure 50 tubes are covered. The [GaDdj@) current TS did not ensure tube coverage. S/l587 WCGS-Conversion Congparison Table-ITS3.4
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.5-3 APPLICABILITY: CA. WC l l REQUEST: CTS 4.4.1.2.2,4.4.1.3.2 and 3.4.1.4.1.b and ITS 3.4.5,6 and 7 (Callaway and l Wolf Creek) Comment: Ten percent wide range level was specified as the necessary heat sink level. Now in the ITS the levelis narrow range. Was this is a known error in the TS that is now l being corrected or was this just discovered as part of the conversion effort? Please I provide the technical basis for concluding that 10% (4% for Callaway) narrow range is adequate. Additionally explain why different narrow range level values are used at each plant and why wide range level is used in Mode 5 at one and not the other. FLOG RESPONSE: The CTS Bases for Reactor Coolant Loops and Coolant Circulation is silent on the background behind the 10% wide range value specified. During the ITS conversion process, industry traveler TSTF-114 was developed (and subsequently approved by NRC), to recognize the importance of keeping the SG tubes covered as discussed in NRC IN 95-35," Degraded Ability of SGs to Remove Decay Heat by Natural Circulation." Since 10% on the wide range level instrumentation does not ensure the SG tubes are fully covered, the values were changed in the conversion amendment. This was not an error known prior to starting the amendment development for ITS. Plant Specific Discussion The basis for using 10% narrow range for Wolf Creek was that NUREG-1431 provided a level Wolf Creek believed was conservative in relation to the CTS. After further review of this comment and Comment Number 3.4.5-2, Wolf Creek believes that it is appropriate to change the value to 6% narrow range since it is used throughout the Emergency Operating Procedures (EMGs), it has operator awareness because of the EMG familiarity, and ensures an SG water level approximately 100 inches
- above the top of the highest SG tube. Additionally, Wolf Creek believes
! that the wide range level instrumentation should be used for MODE 5 l since it is calibrated for cold conditions and provides a larger span. Therefore, the ITS and Bases are revised to reflect the use on 6% narrow range and 66% wide range in MODE 5. ATTACHED PAGES: l See attached pages in the response to Comment Number Q 3.4.5-2. 1
4 ADDITIONAL INFORMATION COVER SHEET 1
\
ADDITIONAL INFORK., TlON NO: O 3.4.6-1 APPLICABILITY: DC, CP, WC, CA l l REQUEST: Difference 3.4-02 l j Comment: The difference states that the STS doesn't cover all possible configurations I and the language of the STS is potentially confusing. Please explain the basis for these comments. FLOG RESPONSE: The STS wording for Condition A, "One required RCS loop inoperable l AND Two RHR loops inoperable", and for Condition B, "One required i I RHR loop inoperable AND Two required RCS loops inoperable", is confusing. This confusion arises from the fact the LCO allows any combination of two RCS or RHR loops, including one RCS loop and one
- RHR loop, to satisfy the OPERABILITY requirement yet Conditions A and j B are worded as if either two RCS loops or two RHR loops, exclusively, l
- were the required loops.
4 By way of illustration, the following scenarios are presented. Assume the LCO's OPERABILITY requirements are satisfied by one RCS loop and , one RHR loop. These loops are serving as the " required" loops. If the RCS loop becomes inoperable, Condition A does not apply because it is
- "ANDED" with "Two RHR loops inoperable" yet one RHR loop remains
- OPERABLE in this scenario. Conversely, if the RHR loop becomes inoperable, Condition B does not apply because it is "ANDED" with "Two required RCS loops inoperable" yet one RCS loop remains OPERABLE.
In fact, the wording of STS Condition B is at odds with the LCO since Condition B requires three loops to be OPERABLE (one RHR and two RCS loops). The FLOG considered this wording to be a potential source of error for plant operators. Since the corresponding CTS specification is not confusing it was adopted in lieu of the STS wording. This confusion also led to the WOG creating a traveler, WOG-109, which was subsequently withdrawn and superseded by TSTF-263 which is currently under NRC review. TSTF-263 presents a very similar approach to that used by the FLOG to correct STS 3.4.6; however, TSTF-263 has not been incorporated by the FLOG. TSTF-263 was not issued until several months after the FLOG submittals. The changes incorporated in ITS 3.4.6 are based on the CTS which has less rigid logic connectors than the STS. ATTACHED PAGES: None
_. . .. -. - .._. - . - - . - _, . - . _ ~ . - . - . _ - - . ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.7-1 APPLICABILITY: WC REQUEST: ITS 3.4.7.2 (Wolf Creek) Comment: It should read " required SGs" rather than " required Sgs". . i FLOG RESPONSE: The smooth copy of the ITS has been marked to read " required SGs." A ' final review of the smooth ITS and ITS Bases is planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases ATTACHED PAGES: None l l 1 1 l l l I i l i l l I 1 l l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.7-2 APPLICABILITY: WC REQUEST: ITS LCO Bases 3.4.7 and 3.4.8 (Wolf Creek) Comment: The TS condition " Loops Not Filled" should be defined in the TS Bases subject to the Bases Control Program ed not in an unnamed plant procedure for which the control mechanism is not specified. FLOG RESPONSE: During the development of the conversion application, the current Technical Specification Clarifications were reviewed to determine if the ; clarification should be incorporated into the ;TS or ITS Bases. Wolf Creek had developed a Technical Specification Clarification to define " loops filled" and " loops not filled" and has subsequently incorporated these definitions into general operating procedures. The added text to the ITS : Bases was to indicate that these definitions are contained in plant procedures. Wolf Creek is revising the ITS Bases to delete the reference to plant procedures. ITS Section 5.4.1 requires written procedures be established, implemented and maintained. Control mechanisms are in place to ensure procedure changes are reviewed and approved. Therefore, the Bases text is unnecessary and the WCGS ITS is now consistent with NUREG-1431, Rev.1. j ATTACHED PAGES: Encl. 58 B 3.4-35, B 3.4-38 l l l l \
RCS Loops-H00E 5, Loops Filled B 3.4.7 BASES J 0 3 A .1 7.1 i - APPLICABILITY In MODE 5 with RCS loops filled (j this LCO requires forcedoncircula;Ms.dE-j or me reactor coolant-if~^t to- . l remove decay heat from the core and to provide proper boron l mixing. One loop of Rm provides sufficient circulation for these purposes. However, one additional RR loop is required to be OPERABLE. or the secondary sideC-i+-; water level of at _ least two SGs it required to be a i F a 3,4g.2. 66 */, 4 5.45 G ' Operation in other MODES is covered by: l l LCO 3.4.4. "RCS Loops-H0 DES 1 and 2": LCO 3.4.5 "RCS Loops-H0DE 3": LCO 3.4.6, "RCS Loops-H0DE 4": , LCO 3.4.8, "RCS Loops-HODE 5. Loops Not Filled": l LCO 3.9.5, " Residual Heat Removal (RHR) and Coolant l l Circulation-High Water Level" (HODE 6); and l LCO 3.9.6, " Residual Heat Removal (RHR) and Coolant ! Circulation-Low Water Level" (HODE 6). ACTIONS A.1 and A.2 . 414 T~1 ps. A.s 3 l .- If one Rm loop is - rable and the required SGs have secondary l W--M W je-) side
- water levels <' redundancy for heat removal is lost.
Action must be initi imediately to restore a second RR loop to OPERABLE status or to restere the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The inmediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal. ! B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Notes 1 andL4, or if no loop is OPERABLE, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. Addition:of[ borated water?with'aiconcentration greaterithan or equal to-theteiniana required;RWST concentration but,lessrthan the , actual:RCSiboron concentration ~shalTYnot be considered:a: reduction 2in: boron concentration.(Ref!T2). To prevent inadvertent criticality duringia boron dilution, forced circulation from.at0MoneJRCP . is required to provide proper mixing! .74 Fu;r4; th; .;.c.rMn te critiality in thi'; tyg of ;Fr;ti;n. The imedia'ce Completion Times reflect the importance of maintaining operation for heat removal. 1 (continued) i WCGS-Mark-sqp ofNUREG-1431-Bases 3.4 B 3.4 35 $/1/587 l l l
.._ _ __m.-. _ _ _ _ _ _ . _ _ _ . - _ . _ . . _ . . _ _ _ _ _ _ . _ . _ _ _ _ . _ . . _ _ _ _ _ . _
l RCS Loops-H0DE 5. Loops Not Filled B 3.4.8 BASES LCO The purpose of this LCO is to require that at least two RHR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via the RIR System unless forced flow is used. A minista of one running RHR pump meets the LCO requirement for one loop in operation. An additional RHR loop is required to be OPERABLE to meet single failure considerations. Note 1 permits all RHR pumps to be d; crer;iz; tremoved_from op}rationTfor s 1"houC15 ;;;inutes den ;; itching fr, ;re 1;;p t; enether. The ciretastances for stopping bott RlR pumps are to be gh limited to situations when the outage time is shorg ano core
/ outlet temperature is maintainedQ)10*F below saturation ja: 24.aoi i l Ltemperatur
, ,,g,a I epermen;eVThe Jan "J:", Lt; prehibit; fer;;d fl;; i::;;re ditt;;n ThelNotelrequires
;t,,,,,cd. er dreini.
p 3,w.uen.g @he rZvessel Water]everbe[above:thOessellflange~to; ensure operatingLRfR:ptmp willinot beLintentionallyw.= w t zed , during mid[ loop operations. 1 Note 2 allows one RHR loop to be inoperable for a period of s 2 hours, provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be i performed on the inoperable loop during the only time when these tests are safe and possible. An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. I Q 3.4772] APPLICABILITY In MODE 5 with loops not filledMgpMartfcadipe3s this LC0 requires core heat removal ana coolant circulation IFy the RHR System. One RHR' loop provides sufficient. capability for this purpose! 'However, one additional RIR: loop;is required to be OPERABLE to meet single failure considerations. Operation in other MODES is covered by: LC0 3.4.4. "RCS Loops-H0 DES 1 and 2": LCO 3.4.5 "RCS Loops-H0DE 3": LC0 3.4.6. "RCS Loops-H0DE 4": LC0 3.4.7. "RCS Loops-H0DE 5. Loops Filled": LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (H0DE 6): and LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (H0DE 6). (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 38 S/1/5/97
! l l ADDITIONAL INFORMATION COVER SHEET l l l [ ADDITIONAL INFORMATION NO: O 3.4.8-1 APPLICABILITY: DC, CP, WC, CA l REQUEST: Difference 3.4-48 ; Comment: It is unclear why TS 3.0.4 would not apply. If this change is to be considered it should be done on a generic basis. FLOG RESPONSE: A Reviewer's Note in STS LCO 3.0.4 states: "LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the ; Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for l entry into a MODE or other specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs to facilitate NRC staff review of a conversion to the STS " Based on this Reviewer's Note, a matrix of this evaluation was placed in the NSHC LS-1 in Enclosure 4 of the Section 3.0 package (Attachment No. 6). JFD 3.4-48 has been revised to incorporate additional justification from NSHC LS-1 in Enclosure 4 of the Section 3.0 package (Attachment No. 6). JFD 3.4-48 has been revised to include:
"LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. ITS LCO 3.4.8 is modified by a Note stating: "While this LCO is not met, entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is ,
not permitted." The transition from MODE 5 (loops filled) to MODE 5 (loops not filled) removes the steam generators as a decay heat removal system while the RHR System is potentially degraded. Therefore, the Note ensures that the transition is precluded if LCO 3.4.7.b (two SGs) were chosen (in lieu of the second RHR loop) to ensure decay heat removal capability prior to draining the RCS." It should be noted that the Applicability Bases for ITS 3.4.8 already provides a similar discussion. l ATTACHED PAGES: Encl.6A 9
CHANGE NUMBER JUSTIFICATION T,,, measurements are not obtainable for a non operating l loop. ; 3.4 47 ITS SR 3.4.11.1 contains a Note which exempts the cycling . of the block valve when it is closed in accordance with Required Actions of Condition B or E of LC0 3.4.11.' ] However. Required Action A also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also , exempt performance when the block valve is closed 1.- I accordance with Required Action A.1 as the block val a should not be opened when the PORV is inoperable. This
- change is consistent with NUREG 1430 and NUREG 1432 in as much as the block valve cycling _is exempted under ._
Conditions A.,B. and E.fS e p yet to WAlocVvalueWu - ~ isytiLinfd in)t6 qui ActMn A. . Note to SR ) Q M khD (3.4.11.1 will be revised to not require the surveillance {Q performanceiftheblockvalve(s)isclosedpePjt 1 on e 'o( 64 LCO . / 'A . ce r the ock alv s) ' r ved 'n R ir Acti s B.2 nd E ,t sur eill ce an t i t. ive he wo ing an " met' to " rf i } .t wordi of .4 .1 re sed o e th[) ate i he C ition and exc tion. D his change is consistent with traveler WOG 87. anear W%31Q 3 4.II-4_] 3.4 48 fAYte is added to ITS 3.4.8 ACTIONS. indicating that]
- entry into MODE 5 Loops Not Filled from MODE 5 Loops %
- Filled is not permitted while LC0 3.4.8 is not met. The i addition of this note is based on the performance of a
' plant specific LC0 3.0.4 matrix (see CN 1 02 LS 1 of
- Q CTS 3/4.0 package). 1 4 M a.T c, A . 9 4 _, 93 WO l 3.4 49 LC0 3.4.12. "[LTOP) System." provides four differenct i
methods for pressure relief. Any of the four may be used. However. Surveillance Requirement 3.4.12.5 requires
- testing whether or not the equipment is being credited to
! meet the LCO. The proposed change adds the word
- " required" to the Surveillance to exempt its performance if the equipment to be tested is not being used to meet the LCO. In addition, two editorial changes were made.
The LC0 requirement presentation was clarified. Also, the-Note to SR 3.4.12.8 was revised to replace " required to be met" with " required to be performed" since the
" performed" nomenclature is appropriate here, consistent wit _h the CTS. This change is consistent with traveler F-28)0 {Q 3.4.12-1 _
WCGS-DifferencesfromNUREG-1431-ITS3.4 9 S/1587
INSERT 6A-9a 0 3.4.8-1 LC0 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall l not be prevented. In addition, LCO 3.0.4 has been revised so that it is only l applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1, 2, 3, and 4. The MODE change restrictions in LC0 l 3.0.4 were previously applicable in all MODES. ITS LC0 3.4.8 is modified by a Note stating: "While this LC0 is not met, entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is not permitted." The transition from MODE 5 (loops filled) to MODE 5 (loops not filled) removes the steam generators as a decay heat removal system while the RHR System is potentially degraded. Therefore, the Note ensures that the transition is precluded if LCO 3.4.7.b (two SGs) were chosen (in lieu of the second RHR loop) to ensure decay heat removal l capability prior to draining the RCS. l l INSERT 6A-9b 0 3.4.11-4 In addition, Notes are added to Conditions C and F stating that these Required ' Actions don't apply when the block valve (s) is inoperable solely as a result of its power being removed per Required Actions B ^ or E.3 as a result of an inoperable PORV(s). In this scenario Condition B ur E is entered as a result of an inoperable PORV(s). If one PORV were inoperable and incapable of being manually cycled Condition B would be entered at time zero, to. Required Actions B.1 and B.2 would close the associated block valve and remove its power within time to + 1 hour. If, as a result of block valve power removal l per Required Action B.2, Condition C were then entered Required Action C.1 would require the associated PORV to be placed in manual control within time to + 2 hours. However, the reason for originally entering Condition B is that the associated PORV is inoperable and can't be manually cycled, thus there is nothing to be gained by placing the PORV in manual control. The PORV inoperability may be such that the PORV can't be placed in manual control (e.g., blown control power fuse), in which case Required Actions C.1 and C.2 can't be met. In addition, Required Action C.2 (restore bico valve to OPERABLE status) can't be satisfied as long as power is removed from the block valve. Restoring the PORV to OPERABLE status within time to + 72 hours allows the plant to exit Condition B. If power were not restored to the block valve at this time, the new Note on Condition C would have no standing and Condition C would be entered. Similar conclusions can be drawn for the relationship between Conditions E and F. If Condition E is the original Condition entered, there is nothing to be gained by Required Action F.1 and Required Action F.2 can't be satisfied with block valve power removed. With F.2 not satisfied. Required Action G.2 would require the plant to be in MODE 3, but Required Action E.4 would have already had the plant in MODE 3 two hours earlier.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.9-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.4.9 Comment: Does 92% (90% for Diablo Canyon) in the pressurizer ensure that upon an inadvertent Si that the pressurizer will not overfill before the operator is assumed to take action? Other plants have lowered this limit (Robinson) or qualified the PORVs for wi.er (Millstone 3). FLOG RESPONSE: ITS Surveillance Requirement 3.4.9.1 requires the pressurizer water level to be less than 92% (90% for Diablo Canyon). This requirement is not related to the assumptions used in the inadvertent safety injection analysis. The basis for this requirement is given in the ITS Bases for SR 3.4.9.1 (as clarified by NRC approved TSTF-162), which states that it is to ensure provision of a minimum space for a steam bubble which is an assumption in the safety analyses (i.e., the pressurizer must not be water solid). This maximum pressurizer levelis not assumed in any safety analysis. ATTACHED PAGES: None
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.9-3 APPLICABILITY: DC, CP, WC REQUEST: Difference 3.4.17 (Wolf Creek, Diablo Canyon and Comanche Peak) Comment: TSTF-93 Rev. 3 was approved with a reviewer's note which says that for non-dedicated safety-related heaters which normally operate the frequency is 18 months and for dedicated safety-related heaters which normally don't operate the frequency is { 92 days. Each of the plants is asking for the 18 month frequency but it is unclear from I the submittals if they meet the criterion. Please provide information demonstrating consistency with the TSTF. FLOG RESPONSE: DCPP and WCGS have two-groups of non-safety related pressurizer backup heaters. The pressurizer heaters, together with the pressurizer spray valves, are used to contro! RCS pressure. For DCPP, the NRC recently approved (6/5/98) changing the CTS SR 4.4.3.2 from 92 day to " Refueling Interval" in L:A 126/124. For Comanche Peak, the pressurizer heaters used to satisfy the pressure control function are comprised of one proportional control group and three backup groups. The design and operation is consistent with the basis for an 18 month surveillance described in Section 6.6 of NUREG-1366 (which i was the basis for TSTF-93). The heater groups are normally connected to the emergency power supplies (two to each Class 1E train of emergency power) and normally operate. CPSES will revise the 3.4.9 BASES to reflect the NUREG-1366 basis for the 18 month frequency. ATTACHED PAGES: Encl. 5A Traveler Status page
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 i TRAVELER # S.TATUS . DIFFERENCE # COMMENTS TSTF-26 Incorporated 3.4-32 Approved by NRC. TSTF-27, Revh incorporated 3.4-33 (Ag-Ub 16QQ14."L-il 5 TSTF-28 incorporated 3.4-22 Approved by NRC. TSTF-54, Rev. I !acorporated NA (Arh N *JM71a. 5.+ep9 I TSTF-60 lacorporated 3.4-15 Approved by NRC. TSTF-61 Not incorporated Minor change that is adequately addressed in the Bases. TSTF-87, Rev.h lacorporated 3.4-31 (khd 63 AJQ /FIE.3.N24- l TSTF-93 Q Incorporated 3.4-17 {Apprevwf 6[UAQ /47.4. iJ l TSTF-94 h Not incorporated NA Retained current TS. [7M 3.4.ce s-l YV2/Inc$rforatKy M IQ 3.4.> t TSTF-108, Rev. I Not Incorporated NA LCO 3.4.19 does not apply. TSTF-113, RevhI Incorporated 3.4-39 q 3,4: .3} TSTF-ll4 incorporated NA Approved by NRC. TSTF-il6, Revhb Incorporated 3.4-36 l 4.3.4. 83-2J TSTF-136 Incorporated NA (AprM k,9 MACD /71R.3.6do?f TSTF 137 incorporated NA [dppe,vuL63 AM g /7A 3.W- Sof I TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6 TSTF 151 h Incorporated NA /TA J.e cof] TSTF-153 Incorporated 3.4-01 hrwv'*6 by A4f.C]/723.9-not / TSTF-162 Incorporated NA (Appraved by AIAC.]/73t.J.4 <20 6,] GliOG4tfBef"D Incorporated h3.4-45%s) See also Cns 3.4-18 and 3.4 20.19 8 4 Il~k_I (WiMp90 '15fFa$ Incorporated 3.4-35 [4 3.4.Il .21 IlheC.69:"Ramro incorporated 3.4-10 DCPP onlyhffbi.4 by ilRCJN3.4 dof I ([WOG-87,'Rev) Incorporated 3.4-47 143.4.13-4) M* Incorporated 3.4-40 Applicable to Callaway and (11sts:-28@ Wolf Creek only. [@ s.4.1-M Incorporated 3.4-49 \43.+.L'2.-1i 5/15M7
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.10-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.4.10 Bases Applicable Safety Analyses Comment: What justifies the differenr.es between the ITS Bases and the STS Bases and between the plant Bases (especially Callaway and Wolf Creek) of the lists of possible over pressurization events? FLOG RESPONSE: These Bases changes reflect each plant's licensing basis as expressed in their respective versions of USAR Chapter 15. Plant Specific Discussion in the Wolf Creek Applicable Safety Analysis Bases for ITS 3.4.10, changes were made to items b and e. The change to item b deleted the Loss of reactor coolant flow and added Feedwater line break as new item b since the analysis of this event shows the primary and secondary side safeties li't, as discussed in USAR Section 15.2.8 (see USAR Figures 15.2-14 and 15.2-19 for the pressurizer pressure and volume transients). The change to item e denotes that this event is not a station blackout; standby power is available from the diesel generators consistent with the discussion in USAR Section 15.2.6. Further review of this section of the ITS Bases determined that additional changes are warranted. The Loss of reactor coolant flow accident is being retained in the Bases since analysis shows that primary and secondary side safeties willlift. Item c is being revised consistent with the discussion in USAR Section 16.2.2 and 15.2.3, i.e., no loss of external load analysis is presented in USAR Section 15.2 since the turbine trip is more limiting. RCCA ejection should be added as new item h since there is pressure surge analysis, discussed on page 15.4-35 of Section 15.4.8.2.2, that is incorporated by reference to WCAP-7588, Rev.1-A, January 1975,"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetic Methods," Risher, D. H., Jr. This pressure surge analysis concludes that with an ejection worth of one dollar at BOL and HFP conditions, the resulting stress levels do not exceed faulted stress limits. All of the changes in this Bases section transform a generic discussion to one that applies to this plant specifically. ATTACHED PAGES: Encl.5B B 3.4-47 1
Pressurizer Safety Valves B 3.4.10 BASES (continued) BACKGROUND The consequences of exceeding the Amcrican Society of (continued) Mechanical Engineers (ASME) pressure limit (Ref.1) could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation. APPLICABLE All accident and safety analyses in the UFSAR (Ref. 2) that SAFETY ANALYSES require safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure. The overpressure protection analysis (Ref. 3) is also based on operation of rJeesafetyvalves. Accidents that could result in overpressuriza lon if not properly tcrainated include:
- reAttne tas.s.oes-\ \
- a. Uncontrolled rod withdrawal from full power:
( u. nae.1e+s.
@ 3.4.in - O
- b.
- Loss if rcoctor c;clont flow kdwater line brea . _
- c. Loss of external electrical load- g34,te : j
- d. Loss of normal feedwater: /
- e. Loss of all non emergency AC power to station auxiliaries:
and
- f. Locked rotor.
. =
e ailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in ricnts c. d. ond c '; ksc) the above events to limit the pressure increase. Compliance with this LC0 is consistent with the design bases and accident analyses assumptions. Pressurizer safety valves satisfy Criterion 3 of the NRC Policy Stat;x nt. 10 CFR 50.36(c)(2)(ii). Red cluster control e.wsmMy s'3ech@ los.4.io- } LC0 The three pressurizer safety valves are set to open at the RCS design pressure (2500 psve 2485 psig), and within the ASME specified tolerance, to avoid excee:' ig the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The upper and lower pressure tolerance limits are based on the i 1% tolerance requirements (Ref.1) for lifting pressures above 1000 psig. (continued) l WCGS-Alark-up ofNUh3G-1431 - Bases 3.4 8 3.4 47 S/V5i97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-1 APPLICABILITY: DC, CP, WC, CA REQUEST: Change 4-04 LG Comment : The requirement is in the CTS and the STS. The justification for not putting it in the ITS is that automatic actuation to open is not required. However, proper calibration also ensures that the PORV does not prematureiv open creating as stated in the Bases "in effect a small break LOCA." FLOG RESPONSE: There are two CTS LCOs (3.4.9.3 (3.4.8.3 for CPSES) & 3.4.4) and corresponding ITS LCOs (3.4.12 & 3.4.11) controlling pressurizer PORV operability. One of these, CTS 3.4.9.3 (3.4.8.3 for CPSES) and corresponding ITS 3.4.12, governs their operability as part of the LTOP/COMS system. Both the CTS (SR 4.4.9.3.1.b or 4.4.8.3.1.b for CPSES) and the ITS (SR 3.4.12.9) reqJire CHANNEL CALIBRATIONS of the LTOP/COMS PORV actuation channels every 18 months to support this function.. The second of these, CTS 3.4.4 and ITS 3.4.11, governs the operability of the PORVs and their block valves as isolable relief valves. While the ability to open the PORVs manually and to isolate a stuck open PORV using its block valve are considered safety-related capabilities, the ability of the PORVs to act as automatic relief valves in Modes 1,2, and 3 is not a safety function in the current licensing basis. The pressurizer safeties fulfill both the RCS Code overpressure protection function and the automatic pressure relief function assumed in the accident analyses. For this reason, STS 3.4.11 does not have a CHANNEL CALIBRATION surveillance requirement. SR 4.4.4.1.b is therefore moved out of the technical specifications by DOC 4-04-LG. This is appropriate since automatic actuation of the PORVs is not a currently credited safety function in Modes 1,2, or 3. Requirements that are not needed to support the safety analyses are moved out of the Technical Specifications, reflecting the philosophy and content of NUREG-1431. The premature opening of a PORV is considered to be a small break LOCA. A LOCA is an unisolable leak or break in the RCS. A stuck open pressurizer safety valve would constitute a LOCA. One of the design functions of the PORVs is, however, to reduce the risk of a stuck open safety by having actuation set points below those of the safeties. As stated in the STS Bases for ITS 3.4. i1, LCO, a stuck open PORV could be isolated by closing its safety related block valve, thus avoiding a LOCA. The automatic actuation of a PORV at a pressure lower than its nominal design set point is not desirable, but is not outside the safety analysis. ATTACHED PAGES: None
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-2 APPLICABILITY: DC, CP, WC, CA REQUEST: Change 4-08 LS 34 and Difference 3.4-35 Comment: WOG-60 has not yet become a TSTF. FLOG RESPONSE: WOG-60 has beers approved by the TSTF and is designated as TSTF-288. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-288 was modified from WOG-60, Rev.1, and these modifications have been incorporated into the ITS (editorial SR Bases nlote change). The FLOG continues to pursue the changes ; proposed by this traveler. l ATTACHED PAGES Encl.3A 8 Eucl. 3B 5 Encl. SA Traveler Status page Encl. SB B 3.4-58 Encl. 6A 7
i i ( l CHANGE l NUMER IGtlC CESCRIPTION b 28D]G 3.4.tl-2.
- 4 08 LS 34 Consistent with traveler the requirement to )
l perform the 92 day surveil ance of the pressurizer PORV l block valves and the 18 month surveillance of the i pressurizer PORVs (i.e., perform one complete cycle of l each valve) is revised to indicate that the surveillance ! is only required to be performed in MODES 1 and 2. This ! is consistent with the recommendations of Generic Letter 90 06, " Resolution of Generic Issue 70, ' Power 0perated Relief Valve and Block Valve Reliability.' and Generic Issue 94, " Additional Low Temperature Overpressure l. Protection for Light Water Reactors,' Pursuant to 10CFR50.54(f)," June 25, 1990. 4 09 LS 36 The requirement to perform the 92 day surveillance of pressurizer PORV block valves (i.e., perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of l the current TS LC0 3.4.4. This change is acceptable l because no credit is taken for the automatic actuation of i the PORV in Modes 1, 2, or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube Rupture (SGTR) accident. However, the capability to manually cycle the PORVs will be unnfected by this change. This change will not affect the ability of the block valve to open, if closed to meet ACTION a, in the ! mitigation of an SGTR. Deferral of the block valve ( cycling surveillance will not diminish the design
- capability of the block valve to open against differential j pressures that would be present after an SGTR since the i block valves are capable of opening again.st 2485 psig, the
! safety valve lift pressure, whereas pressurizer pressure l decreases after an SGTR. This change is consistent with traveler WOG 87. @ sEh M e W. (cp s.4.n-4. l 5 01 A This change moves the steam generator tube surveillances l to ITS SR 3.4.13.2 and Administrative Controls Sections
- j. 5.5.9, " Steam Generator (SG) Tube Surveillance Program" and 5.6.10. " Steam Generator Tube Inspection Report."
5 02 A LCO 3.4.5 is deleted for consistency with NUREG 1431 Rev.
- 1. Steam generator operability requirements in MODES 1 4 o are specified in the RCS loop and operational leakage j specifications.
f a WCGS-Description of Changes to CTS 3M.4 8 S/258 7
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 5 of 13 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON C0HANCHE PEAK WOLF CREEK CALLAWAY 4-04 This change moves the requirement to perform channel Yes-Moved to FSAR. Yes- Moved to TRM. Yes- Moved to USAR. Yes-Moved to FSAR. LG calibration of PORY actuation instrumentation to a licensee controlled document. 4-05 The shutdown requirement of CTS 3.4.4 would require the Yes Yes Yes Yes LS-31 plant to reduce To to <500T within 12 hours, rather than go to MODE 4. to address the concern of entering [LTOP] LCO Applicability with inoperable PORVs. For consistency the shutdown requirements of CTS LCO [3.4.8'3 would be similarly revised. 4 06 This change provides a 72 hour completion time to restore an Yes No - Already part No - Already part No - Already part LS-32 inoperable block valve. with the PORY placed in manual of CTS. of CTS. of CTS. control mode. The current TS requires the block valve to be restored within one hour, or remove power from the solenoid. 4-07 This change provides a two hour completion time for Yes No - Already part No - Already part No - Already part LS-33 restoring an incperable block valve when more than one block of CTS. of CTS. of CTS. valve is inoperable and 72 hours to restore the remaining valves. The current TS requires the block valve to be restored within one hour for one or more valves inoperable. the requirement to perform Yes Yes Yes 4-08 Consistent with traveler Yes LS-34 the PORY and block valve cycling surveillances is revised pg such that the surveillance is only required to be performed in MODES I and 2. @ a.4.W 2. [ Consistent with traveler WOG-87, the requirement to perform Yes Yes Yes Yes 4-09 LS-36 the 92 day surveillance of the pressurinr block valves (i.e.. perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of the current TS LCO 3.4.4. 4 3
~
This change moves the Steam Generator Tube Surveillances to Yes No - Same as CPSES Yes Yes 5-01 A ITS SR 3.4.13.2 and the Achinistrative Controls Sections change 1-14-A for 5.5.9 and 5.6.10. CTS Section 3/4.0. inadehkn, a.uote 4 added b Actidn@o prevent enivj sot
- M 3^+'H-4 l duc.to shopemb4a. PORVM -
urvier Ac-tio'ndb erc.} N (. A P93.A Py[d eg [A peSyg sp pp4* A SG YA k fA , I[M M f(h7
. - - . . - . ..- . . - - - - . - . - . ~ - . - . - . - - . . . - . -
l l 1 INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 ! I . TRAVELER # STATUS . DIFFERENCE # COMMENTS 1 TSTF-26 incorporated 3.4-32 Approved by NRC. TSTF-27, Revh locorporated 3.4-33 M . 3 byllR Q Q 3.14."L-il TSTF-28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev.1 Incorporated NA { AP p g by M M me..t+ mog l > TSTF-60 Incorporated 3.4-15 Approved by NRC. I 1 TSTF-61 Not Incorporated Minor change that is adequately addressed in the j Bases. TSTF-87, Rev.h (dhd bg AQ /718.3.49MW j
~
incorporated 3.4-31 TSTF-93 Q incorporated 3.4 17 {Apprewd63 0 /97.4. iJ] TSTF-94 N Not Incorporated NA Retained current TS. [7'M 3.v-de s-[
- [TSJrTVff_InierEorat p r __ / e l4 3.4 9 L TSTF 108, Rev. I Not Incorporated NA LCO 3.4.19 does not apply.
TSTF-Il3, RevhI Incorporated 3.4-39 4 3.4.11- 33 , TSTF-ll4 Incorporated NA Approved by NRC. TSTF-ll6, Rev[' Incorporated 3.4-36 j 4.3.4. 83-2J . m , j TSTF-136 incorporated NA [Aymys4. kg MAC] /TIR.3.6det f
- TSTF 137 Incorporated NA [4pprevsd 6,y MWJ/74 3.#- Sof l TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6 TSTF-151 Q Incorporated NA /7"4 J. W j
- TSTF-153 Incorporated 3.4-01 hrwved,byMED/TR3.f-not /
TSTF-162 Incorporated NA (App aved by NK.)/738.#.4-#G} l GIMGWD See also Cns 3.4-18 and 3.4-20.19 8 4 IL*L T Incorporated h3.4-45%5)2. ( *15fFa@ Incorporated 3.4-35 [q 3.461 2 \ , EheG.6P'RawO Incorporated 3.4-10 DCPP onl int NEE.")MI.Nf I (WOU-87,'Rev3 Incorporated 3.4-47 [ 4 3. 4.81- 4 ) M* Incorporated 3.4-40 Applicable to Callaway and i (Ters::=23Q Wolf Creek only. [@3 4.H2.) Incorporated 3.4-49 \Q.3.+.I'2.-I $ 4 S/15/97
---yw. - , - - -.,*.w -
=+7
l l 1 Pressurizer PORVs B 3.4.11 BASES (continued) SURVEILLANCE SR 3.4.11.1 (continued) REQUIREMENTS _ Nyg,4 E is+__ pable- ?bei ma lly-cyc. .t maxi Com) ion T1 o res re t PORV a ope the bl k va e is (7p urs, ch is 1 wi in the lowa e limi (25 to {4xtend block lve F equency f 92 ys. F the e. t se I test uir s wou be c eted the r peni of a r ntly c ed bl valve nr oratio of t PORV ERABL tatus ( .e. e letio of the uir Act (fulfi s the SR1 Qg ,. gg g h yeg cs4.ti-2.} 4 N Be Notel modifies this SR by stating that it is not required to be met perfomed.with the block valve closed, in accordance with the Required Actions of this LCO. Notei2; modifies;this SR?to allow!entryJintoiand: operation it ~3fprior:to; performing the SRUThis; allows:the: test to be: rformed fin'H0DET3!under operatingitemperature and press e conditions,; prior lto entering MODE (17or'27 % ggggg% p3.4.sl4l condiW n i A c. reno -the. M *b umiSosabic. lesa fr m me. #Cc5 siht.a. SR 3.4.11.2 % pon.v.3 33rea.g .;,*Perabic . _ SR 3.4.11.2 requires a complete cycle of each PORV. Operating a l PORV through one complete cycle ensures that the_PORV can be i manually actuated for mitigation of an SGTR.fTpe Frepency ofM*A'# I , 8 gorfths f1basepon a typfcal refDeli(cycM andAndustrd j acfepted4ractiee./The Note modifies tT s SR to-allow entry into
! and_ operation in MODE (3 prior to performing the'SR. This_ allows the test:to be performed in MODE ~ 3 under operating temperature
- and: pressure. conditions, prior to entering. MODE 1 or.2.~ In .
, accordance:with Reference 5' administrative controls require this test be:perfomed-in MODE 3:or 4 to adequately: simulate operating j AemperatureandpressureeffectsonPORVoperation. _ i e a+xDrifnce. haslaserviu sheenTc++vh3 &,A these vaM.s usustl9 . pass tu. sum 5b_ a *** P* erb
- ok# the requuhr4 Wey,m sequen %e Frequaa b es.cwansw.aa. h o ra tio.s t;q .-t po mt . -
5"s 3A.11.3 Oper; ting th; ;;lcreid eir ;;ntrol volv;; ;nd ; tad velv;; en th;
;ir ;;;u;;leter; ca;;rc; the P0P," ;;ntrol ;y;tc; ntu;tc;
. Preperly When ;;11;-d spen. Th; Tr;qa rcy ;f [10] renth; i; b ;;d on ; twical refueling cyci; and th Pr;quercy of th; other Surveillen;;; u;;d t; der.~n;tr;t; P0"" OPE"J"!LI"I C" 2.4.11.4
";i; Surv;illera; i; not required for pl=t; With p;r;;rint IE p;.;;r ;upplic; t; the volv;;.
(continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 8 3.4 58 S/VS/97
CHANGE NUMBER JUSTIFICATION I purposes (per Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR Noted exception) would not allow the SR to be considered to be met until the appropriate conditions were available for it to be performed without entering the actions. The Note to these SRs would allow startup in Mode 3 if the SR had not been performed during the l required frequency, but would limit the exception to prior l to entering Mode 2. The change is consistent with traveler TsTF -2% cp.:s.4.n . 2 ] l 3.4 36 SR 3.4.13.1 and LCO 3.4.15 are revised per traveler TSTF- l 116. The note addresses the concern that an RCS water inventory balance connot be meaningfully performed unless the unit is operating at or near steady state conditions. The note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only b: allowed to be deferred for 12 hours after re establis ting steady state conditions. 3.4 37 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 3,g g 3.4 38 _ f----P'"Consi s nt hT 105 the d ails o themet)6d y - . g , whi th CS f rat are v rified removerfr S
/ **.
N *0]PP "". N"
. 1. to t Bas , al Bases Movin this i formati it t s the se of p cisio eat b nc l 'QMD57-ps, d other accepta e met s in o er p fo j thJ verifi - ion and s cons tent w ht N EG 431 philosop of movi clari ng inf rma npd ive detai s out the i to e Bases.
Qeser ,
- 3.4 39 The shutdown requirements of ITS 3.4.11 would require the l plant to reduce T.,, to <500 F within 12 hours rather than l H0DE 4. to address the concern of entering [LT0P] LC0 W14.u-g 3.4.12 Applicability with inoperable PORVs., For
- gQfg>
A consistency, the shutdown requirements of ITS 3.4.16 are
- also revised to allow 12 hours to reduce T.,, to <500 F.
This change is consistent with 3.4 40 Consistent with traveler . . fhe Note to SR 3.4.1.4 {Q3.4.1-2 l would be modified to provi e additional time to perform an ! RCS precision flow rate measurement. The time allowed would be changed from 2? hours to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3 RCS total flow meters) and additional time normally would be required to establish WCGS-Differencesfrom NUREG 1431 - ITS 3.4 7 S/15/97
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.11-3 APPLICABILITY: DC, CP, WC, CA REQUEST: Change 4-05 LS 31 and Difference 3.4-39 Comment: TSTF-113 (presently Rev. 4) has not yet been approved by the NRC staff. FLOG RESPONSE: TSTF-113 Rev. 4 revises the shutdown requirements of ITS 3.4.11 to l allow the plant to reduce T, to <500 F within 12 hours, rather than MODE 4, to address the concern of entering LCO 3.4.12 Applicability with one or more inoperable PORVs. The shutdown requirements of ITS l 3.4.16 are also revised, for consistency, to allow 12 hours to reduce Tm l to < 500 F. ITS 3.4.11 Condition B and C Bases changes have been made to the Callaway submittal to reflect Rev. 4 of the traveler; no changes are required for any other plants'submittals. The FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Encl. 5A Traveler Status page
i t 1 ! INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 '1 TRAVELER # STATUS . DIFFERENCE # COMMENTS TSTF-26 Incorporated 3.4-32 Approved by NRC. TSTF-27, Revh incorporated 3.4-33 b _.3 b 14D@ 5 3'4#3 'll j TSTF-28 incorporated 3.4-22 Approved by NRC. i' TSTF-54, Rev. I lacorporated NA (APP _Q N *JMTE.J.*#9l TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not Incorporated Minor change that is adequately addressed in the Bases. TSTF-87, Rev.h incorporated 3.4-31 (k h d by A; Q /rie.3.4924. ] TSTF-93 Q Incorporated 3.4-17 {Apreeved.6h dR Q /$.7.4. B .3 } TSTF-94 h Not incorporated NA Retained current TS. [TM 3.y.<erf [TS,TFIVff-_inMraty_ y J2C !Q 3.4.b i1 TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply. TSTF-ll3, RevhI lncorporated 3.4-39 4 3.4ll-3] TSTF-Il4 - Incorporated NA Approved by NRC. TSTF-il6, Rev[' Incorporated 3.4-36 1 4.3.4. 83-ad TSTF-136 Incorporated NA {Af rwei, hg hACD /TIR.3.9doif TSTF-137 Incorporated NA [AppM 63 M E J /7A 3.W- Set l TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6 TSTF-151 h Incorporated NA /T4 J.WWWfl TSTF-153 Incorporated 3.4-01 hrwe4 by A14]/743.p-not / l i TSTF-162 Incorporated NA (Appraved by #E.)/73f.# #~ 80(=} 0 EGG Male (D 84il* M Incorporated h3.4-45%r]J. See also Cns 3.418 and 3.4 20.19 l (Wdd90"IWF$ Incorporated 3.4-35 [4 3.4n1.L\ l BNOG49:"Rawo incorporated 3.4-10 DCPP onlhhf d by NEEDN3.V dof I ([WbG-87[Rev3 Incorporated 3.4-47 { 4 3 A.Il-4 ) M" Incorporated 3.4-40 Applicable to Callaway and (Trrf:- 28@ Wolf Creek only. [@ 3.4 1-2.) Incorporated 3.4-49 \ 4 3.+.l 2.- l 5/158 7
ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: Q 3.4.11-4 APPLICABILITY: DC, CP, WC, CA REQUEST: Change 4-09 LS-36, Difference 3.4-47, Change 3-04 and Difference 3.4-31 Comment: WOG-87 has not yet become a TSTF. l FLOG RESPONSE: As discussed during a telecon with NRC Staff on July 30,1998, the above l references to DOC 3-04 and JFD 3.4-31 apply to NRC-approved traveler TSTF-87 and were not intended to be questioned here. Additional changes have recently been added per Revision 2 of WOG-87 and are included in the attached pages below. The addition of the Note to the block valve Action Statement is considered to be administrative in nature as it reflects current plant practice. WOG-87, Revision 2, has been approved by the TSTF group and is expected to be submitted to the NRC expeditiously. Given the nature of the Notes added to the PORV block valve Required Actions and Surveillance Requirement, the FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Encl. 2 4-10 Encl. 3A 8 Encl. 3B 5 Encl. 4 68,69 Encl. SA Traveler Status page, 3.4-25,3.4-26,3.4-28 Encl. 58 B 3.4-55, B 3.4-57, B 3.4-58 Encl. 6A 9 ! Encl. 6B 7 i \ l l
l REACTOR COOLANT SYSTEM l j 3/4 4 4 RELIEF VALVES , 1 l LIMITING CONDITION FOR OPERATION 3.4.4 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1,2. and 3.* l [ NOTE: Seoarate condition entry is allowed for each valve] 4-01-LS 5 - l 6QT,lpN: l a. With one or both PORVs inoperable t:- '- Of :::: : :rr':; 4 02-LS l [and capable of beina manually cycledwithin i hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwisefcontinue restoration ) 4-05-LS , (activities aruDbe in at least HOT STANDBY within the next 6 hours andteduce Tm to <500*@iM4GT-SHWTDOWN-within the following 6 hours.
- b. With one PORV l6esei t:_ 6: t: -- r Ch::'h:n -- :::: x rr' 402-13 l
'r':;:5nd not capable of being manually cycled]within 1 hour either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within thMei6ewr.g-72 hours oricantinue restoration) 4-05-LS 31 :
Factivltes andlbe in HOT STANDBY within the next 6 hours anc(reduce Tm to) 403.M l gb- MOT SHL'TDO'.^^' within the following 6 hours.
- c. With both PORVs inoperable te t --r r c'h::'h , :~:::: x r ' 4-02-LS-6
'- 'z;:Pul not capable of being manually cycled)within i hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve andkontinue restoration]
[ activities andlbe in HOT STANDBY within the next 6 hours N I andFeduce Tm to <500%.. .. . _. ._ . _ _ . . within the following 6 hours. 4 03-M
- d. With one or both block valves inoperab in 1 hour restore the block valve (s) to OPERABLE status or place its associated PORV(s) in manual control Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any l remaining inoperable block valve to OPERABLE status within 72 hours; l
otherwiseicontinue restoration activ ties andlbe in at least HOT STANDBY within the next 6 hours and reduce Tm to <500'F)iM4GT 4-05-LS -SHWTQQWM-within the following 6 hours. l e. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS n 4.4.4.1 ' :f f : t:!M ::-"2 :xx' :^'o additional requirements otherJ 4-04-LG . (than chose required bdSpecification 4.0.5{#jr? OPV l
- ht" 5: Cr c.._.c d C"C^ 9LE et 'r ' xx ;:: *
- xx'h: t"l pc' :.--'n;; :
4-08-LS-34 C"^"'::L C.^ L!S"^7!ON ef'he r-a 't 5:t zrn:- 4.4. Each block valve shall be demonstrated OPERABLE at least once per 92 4-08-LS-34 day y operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTIONQb. or c. 4 09-LS-36 in Specification 3.4.4.
*With all RCS cold leg temperatures above 368'F.
# Only required to be performed in MODES 1 and 2. 4 08 LS-34 >
WOLF CREEK - UNIT 1 -- -- w 4 in_ -- Amendment No. 61 --
- Action cL dess not opply when the blacis vanwM is inspersh solel3 *M
- 8 E*3M r*** ef complyins wib
~ Ac % u b erc.. ~
^
^
14 a .+.ii-+ 1 Mark-up ofCTS 3/4.4 - - - 5/1587
CHANGE NUMBER EMiG DESCRIPTION F 7. h I O 3 4 II~ 1d 4 08 LS 34 Consistent with traveler the requirement to perform the 92 day surveil ance of the pressurizer PORV block valves' and the 18 month surveillance of the pressurizer PORVs (i.e., perform one complete cycle of each valve) is revised to indicate that the surveillance is only required to be performed in H0 DES 1 and 2. This is consistent with the recommendations of Generic Letter 90 06, " Resolution of Generic Issue 70, ' Power-0perated Relief Valve and Block Valve Reliability.' and Generic Issue 94, " Additional Low Temperature Overpressure Protection for Light Water Reactors,' Pursuant to 10CFR50.54(f)," June 25, 1990. 4 09 LS 36 The requirement to perform the 92 day surveillance of pressurizer PORV block valves (i.e., perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of the current TS LC0 3.4.4. This change is acceptable because no credit is taken for the automatic actuation of the PORV in Modes 1, 2, or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube Rupture (SGTR) accident. However, the capability to manually cycle the PORVs will be unaffected by this change. This change will not affect the ability of the block valve to open, if closed to meet ACTION a, in the mitigation of an SGTR. Deferral of the block valve cycling surveillance will not diminish the design capability of the block valve to open against differential pressures that would be present after an SGTR since the block valves are capable of opening against 2485 psig, the safety valve lift pressure, whereas pressurizer pressure decreases after an SGTR. This change is consistent with traveler WOG 87. DsE 3A e W- (ce s.4.n l 5 01 A This change moves the steam generator tube surveillances to ITS SR 3.4.13.2 and Administrative Controls Sections 5.5.9, " Steam Generator (SG) Tube Surveillance Program" and 5.6.10. " Steam Generator Tube Inspection Report." l 5 02 A LCO 3.4.5 is deleted for consistency with NUREG 1431 Rev. l 1. Steam generator operability requirements in MODES 1 4 l are specified in the RCS loop and operational leakage specifications. WCGS-Description of Changes to CTS 3H.4 8 SM587 l
INSERT 3A-8a 0 3.4.11-4 ! , In addition, a Note is added to ACTION [d] stating that it does not apply when , the block valve (s) are inoperable solely as a result of its power being removed per ACTIONS [b or c] as a result of an inoperable PORV(s). In this scenario ACTION [b or c) is entered as a result of an inoperable PORV(s). If one PORV were inoperable and incapable of being manually cycled (per the change discussed under DOC 4-02-LS-6), ACTION [b] would be entered at time l zero, to. ACTION [b] would close the associated block valve and remove its l power within time to + 1 hour. If, as a result of block valve power removal per ACTION [b], ACTION [d] were then entered, ACTION [d] would require the associated PORV to be placed in manual control within time to + 2 hours. However, the reason for originally entering ACTION [b] is that the associated PORV is inoperable and can't be manually cycled, thus there is nothing to be gained by placing the PORV in manual control. The PORV inoperability may be such that the PORV can't be placed in manual control (e.g., blown control power fuse), in which case neither this portion of ACTION [d] nor block valve restoration can be met. In addition, the portion of ACTION [d] requiring block valve restoration can't be satisfied as long as power is removed from the block valve. Restoring the PORV to OPERABLE status within time to + 72 hours allows the plant to exit ACTION [b]. If power were not restored to the block valve at this time, the new Note on ACTION [d] would have no standing and ACTION [d] would be entered. Similar conclusions can be drawn for the relationship between ACTIONS [c and d]. If ACTION [c] is the original ACTION entered, there is nothing to be gained by placing both PORVs in manual control and the block valves can't be restored with their power removed. With ACTION 4 [d] not satisfied, the plant must go to MODE 3, but ACTION [c] would have already had the plant in MODE 3 two hours earlier. Therefore, there is no compensatory action associated with cascading to the block valve ACTION [d] when the sole inoperability-is with the PORV(s). j
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 ' Page 5 of 13 TECH SPEC CHANGE APPLICABILITY NUPEER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY l 4 04 This change moves the requirement to perform channel Yes-Moved to FSAR. Yes- Moved to TRM. Yes- Moved to USAR. Yes-Moved to FSAR. LG calibration of PORY actuation instrumentation to a licensee controlled document. 4-05 The shutdown requirement of CTS 3.4.4 would require the Yes Yes Yes Yes LS-31 plant to reduce T , to <500*F within 12 hours. rather than go to MODE 4. to address the concern of entering [LTOP] LCO Applicability with inoperable PORVs. For consistency the shutdown requirements of CTS LCO [3.4.8] would be similarly revised. 4-% This change provides a 72 hour completion time to restore an Yes No - Already part No - Already part No - Already part inoperable block valve. with the PORY placed in manual of CTS. of CTS. of CTS. LS-32 control mode. The current TS requires the block valve to be i restored within one hour. or remove power from the solenoid, 4 07 This change pruvides a two hour completion time for Yes No - Already part No - Already part No - Already part restoring an inoperable block valve when more than one block of CTS. of CTS. of CTS. LS-33 valve is inoperable and 72 hours to restore the remaining ' valves. The current TS requires the block valve to be restored within one hour for one or more valves inoperable. Consistent with traveler WOG 60. the requirement to perform Yes Yes Yes Yes 4-08 LS-34 the PORV and block valve cycling surveillances is revised such that the surveillance is only required to be performed in MODES 1 and 2. Consistent with traveler WOG-87. the requirement to perform Yes Yes Yes Yes 4 09 I LS-36 the 92 day surveillance of the pressurizer block valves i (i.e. perform one couplete cycle of each block valve) is revised such that it is not required if the block valve is 4 ; closed to meet ACTION a of the current TS LCO 3.4.4. 3 i No - Same as CPSES Yes Yes 5-01 This change moves the Steam Generator Tube Surveillances to Yes A ITS SR 3.4.13.2 and the Acknnistrative Controls Sections change 1-14-A for 5.5.9 and 5.6.10. CTS Section 3/4.0. l In additt'n, e a. klote h Added b Actt6n%+o Pre e entv3 soleb % 3.tll--4 [ duc.4o shopembta. PORYfs) urwier Actions or c ; n ,_ _ n F.c e c - . v . .. .: .. v .eu. r~ w pro a 57;5,9 7
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS
'NSHC LS 36 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS Consistent with traveler WOG 87, the requirement to perform the 92 day surveillance of pressurizer PORV block valves (i.e. perform one complete cycle of each block valve) is revised such that it is not reauired_if the block valve is closed _ to meet ACTION a_ of the current TS_LC0 3.4.4. fin adMMn, a. e46te. is eutdedjt Acien j (fo prevent entr3 solsty du To @eperabje. PORN (sd um4er AC.Tiouf4b er t _g This proposed 15 change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:
"The Comission may make a final determination, pursuant to the procedures in 50.91. that a proposed amendment to an operating 1icense for a faci 1ity licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2 Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety. "
The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change adds a relaxation to the surveillance associated with the
. pressurizer PORV block valves. The quarterly valve cycling will no longer be required if the block valve is closed per any ACTION of the LCO. No credit is taken for the automatic actuation of the PORV in Modes 1, 2. or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube Rupture (SGTR) accident. However, the capability to manually cycle the PORVs will be unaffected by l this change. This change will not affect the ability of the block valve to open, if I closed to meet ACTION a. in the mitigation of an SGTR. Deferral of the block valve cycling surveillance will not diminish the design capability of the block valve to open against differential pressures that would be present after an SGTR since the WCGS-NSHCs-CTS 3M.4 68 5/15/97
l V. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 36 (continued) block valves are capable of opening against 2485 psig, the safety valve lift pressure, whereas pressurizer pressure decreases after an SGTR [ ]. The lack of quarterly block valve cycling, which could extend to a complete cycle since ACTION a allows continued operation with the block valves closed, does not decrease the l likelihood of successful pressurizer relief since power remains available to the l block valve motor operator (s) and the surveillance frequency for the PORVs can be as long as 18 months (tested during each cold shutdown per the IST plan). Quarterly cycling could make PORV seat leakage worse; if the block vcive were to subsequently be unable to close, this surveillance could unnecessarily challenge RCS and PRT l integrity. ,Therefore, the proposed change does not involve a significant increase i in the proiability or consequences of an accident previously evaluated. j l L. G u s e st.T i.s . 3 C ) I Q 3.4.H-41 j
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
l The only acciidents that are potentially associated with this proposed change are those related to a loss of pressurizer relief function. This change does not introduce any new overpressure accidents and the exhting analyses remain valid. Thus, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not affect the acceptance criteria for any analyzed event. The automatic actuation of the PORVs is not credited in the accident analyses for Modes 1, 2, or 3. The PORVs will remain capable of being manually cycled. The margin of safety established by the LCOs also remains unchanged. Thus there is no reduction in the margin of safety from that previously established.
]
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with , NSHC "LS 36" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c); and accordingly, a no significant hazards consideration finding is justified. \ l WCGS-NSHCs-CTS 3M.4 69 S/158 7
INSERT LS-36 0 3.4.11-4 The' addition of the Note to ACTION [d] stating that it does not apply when the ' block valve (s)' is inoperable solely as a result of power being removed per ACTIONS [b or.c] as a result of-an inoperable PORV(s) eliminates operator j ~ distraction caused by the required performance of activities with no safety j benefit. This change has no effect on the successful mitigation of an SGTR l since.the initial premise is that the PORV(s) is unavailable. Elimination of operator actions that have no safety benefit result in an overall benefit to plant safety. 1 l 1 i l l l l j I l l 1 I
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-26 incorporated 3.4-32 Approved by NRC. TSTF 27, Revh Incorporated 3.4-33 (AMb 00DN 5 *DL'Il l TSTF-28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev.1 Incorporated NA (Ah_Q g- h *J M me.3,+isp9 l TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not Incorporated Minor change that is l adequately addressed in the l Bases. TSTF-87, Rev.h lacorporated 3.4-31 (dpp vv.A 6 Al /f1E.3.&24. l TSTF-93 Q Incorporated 3.4-17 {A p p r e d h U M / 9 3.4. i-3] ) l TSTF-94 h Not incorporated NA Retained current TS. [TM 3.v-des f [TSf07 fInMraK _
/ Ml 4 3.4. l" l l TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply.
TSTF-ll3, RevhI Incorporated 3.4-39 4 3.4,18-31 TSTF-ll4 Incorporated NA Approved by NRC. l TSTF-Il6, Rev[' Incorporated 3.4-36 l 4.3 4. 4-3.] TSTF-136 incorporated NA (APPM63 M A C ] / pt.3.6det f TSTF-137 Incorporated NA [4 ppm 69 Mtd)"/7A 3.#- Soff TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements ofITS 3.4.5 and 3.4.6 TSTF-151 Q incorporated NA /TA J 44pfl TSTF-153 Incorporated 3.4-01 hh by MdC]/TRJ.9-not / TSTF-162 Incorporated NA (AppI,ved by NAC.jf3f.J.4 .28 6]
'GilGG4tfBef*D incorporated 3.4-4 % s ) See also Cns 3.4-18 and 3.4 20.R8 4 IL*L) _
[WfMp9o "151F"$ Incorporated 3.4-35 [ 4 3.4.nl . t l IkOGM*RawD incorporated 3.4-10 DCPP onthjjA9 6[slEE"")NJ.Mf I l l ( WOG-87[Rev) Incorporated 3.4-47 l_4 3All-4 ) M* Incorporated 3.4-40 Applicable to Callaway and
, (Tets:-2sg Wolf Creek only. [dps.4 5-2.]
Incorporated 3.4-49 143.+.I'2.-l l S/15/97
pR guded--Ac.tsbe wts. -- - elend *Pfb whin b'*'* V#d" " i"*P"*bt'" Pressurizer PORVs k j*ddg (([c.+ia s.- fk3. 3.4.11 _ 1 ACTIONS (continued) I l COM)ITION REQUIRED ACTION COMPLETION TIME L--- -P 3.4-+l C. One block valve C.1 Place associated PORV 1 hour { q s.4.s t-4 } inoperable. in manual control. M C.2 Restore block valve to 72 hours OPERABLE status. I I i l D. Required Action and D:1 Initiatelactionito Imediate]y [g$g4(agsg associated Completion restor.eiPORV(sFand Time of Condition A. B. block valveltoiOPERABLE ! or C not met. status. ' M
- D.12 Be in MODE 3. 6 hours M
D.t3 Sc ir.."00: 4. Reduce 12 hours Tm to <500!F. (continued) l l WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 25 S/158 7
PrGssurizer PORVs 3.4.11 ; ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
)
Two [;r thr;;] PORVs $23!4!39Mi E. E.1 Initiate;actionJto Immediately 3 inoperable and not rest.oregmeMito g6 - -
- l capable of being OPERABLEjst_atus; manually cycled.
M E.12 Close associated block 1 hour l valves. l l M ' E.23 Remove power from i hour associated block valves. M E.34 Be in H0DE 3. 6 hours M E.45 Sc ir. "00: 4. Reduce 12 hours $83!4e39Mi T,,, to <500*F. F. More than one block F.1 Place associated PORVs I hour valve inoperable. in manual control. M (continued)
------M------ 3.4 4 g Regmred Ac.hcwv. A not app when block. valw n, shoperab [q 3,4,si.4
.soie.13 a.i. a. resuW 4 < ovnpy n3 tah Reqw' red Ac.%ns 3.z.< e. 3.
h ~~ WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 26 S/158 7
Pressurizer PORVs 3.4.11 \ l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 - - -
--- NOTES -- - - -- - -
l
- 1. Not required to be met performed with block valve closed in accordance with the ggg h 2.'
Required Action)of s,elidM'IDrf M 5%Ak Onl D equfredJtoi e;performedJ b n!M00E571 I 434 I gggggs L and'2: Perform a complete cycle of each_ block valveg 92 days gggg ex9pptym. _L .3a e c in
- ~05 I u ; ___'jt ,_1 on of ) 43.4.H-4) tio ~. _or P
.. _ . . . . . . . . . .. ..u. .... .. N0TE . ------- ----- ----- --
. . ~ . . . . . . . - .
Ti.accoraw.a. wm sa, ( 3.4 53 Only required'to be~ performed in' MODES ~1 and 2. iwu gggg:3ES
.= ..:;..:..=x..; = = ;..... = ..............
yg
^ -
SR 3.4.11.2 Perform a complete cycle of each PORV. N N$58% L4C 3.4-4FIj
.PM,
. .. ,J. 1 1. . 83 85 M-
. , , X,v- . . -. ----1 ov. y u A.s_. y. ,..1.,. #.
...mL.
dv, 1---
. 2.J P189
-.AL-6.v; . . ... y
'$:[q,[
J ---A1 ..1... -J L-L ..1... . AL. u33 vvi s k s y a u ns u w w wi ggs vi vw wn vuaww was wtrb I J .- ---.-.1 A--- J .- M#%f%lf -- 1 1 -.. A--- U55 WkVtmas%s a w wvi ea uua u vtsy hviab vi dJ e7 bbvrid e l 1 PM 85 J 1 J i f - J 2. . MfV%11- -J L1- L ..-1.--- --- --- L1. 89 -- LL- -
'$:k.
Wun W . 7 . 1. 7 sht 5 EJ 4 VT)v a wi s%s assywn uwswwd us w buywwik F L 1.v J svywsswasd - *
.2 L J ._ , ..--J 2--- --------. --..--
V5 ass a p ry yvvyw a wwu a a vvvi w e.ww a yw s sw y yvwww a "vv 2 CC3. l t i I l l l t WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 28 5/1587
i Pressurizer PORVs B 3.4.11 BASES ACTIONS B.1. B.2. and B.3 e lo n cm-0 (continued) -- If oneI PORV l is inoperable and not capable of being manually cycled, it must be either restored or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situdon. If the inoperable valve cannot be restored to OPERABLE sta' o it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this additional time, the plant must be brought to ; "00E in which the LCO decs not opply at lea.st' MODE?3 with T.,, < 500*F. as required by Condition D. C.1 and C.2 If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of I hour or place the associated PORY in manual control. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORY remains OPERABLE. the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition B. since the PORVs are net may not be.. capable of mitigating an ;ecrpressure event when pi;ced in ;;nual centrol if the inoperable. block valve is not full open. If the block valve is restored within the Completion Time of 72 hours, the power will be restored and to the PORV. rest; rcd t; 0"C"A"LE status. If it cannot be restored within this additional time, the plant must be brought tc : ",00C in which the LC0 does not ;pply at'least MODE 3'with T.,,$500*F. as required by Condition D. k5ehr [5A 5) I 4 0' 4 ' '4 l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 55 S/1/5/97
, _ . _ . _ _ _ _ . . ~ . . . _ _ _ . .
1 i
)
INSERT B 3;4-55 0 3.4.11-4 The Required A'ctions are modified.by a Note stating that the Required Actions do not apply.if the sole reason for the block valve being declared inoperable is as a ' result of power being removed to comply. with other Required Actions. i
. In this event, the; Required Actions for inoperable PORV(s) (which require the .!
block. valve power to be removed once it is closed) are adequate to. address the condition.
'l l
i. i
)
i l I l k
Pressurizer PORVs B 3.4.11 BASES (continued) ACTIONS 57andT6?(with4he" reactor:vesserhead on) me+nte+nig automatic
~
(continued) P01(FOPERABILITY may be rehiiired." ~'Ses i.00 3.4.12. F.1. and F.2. d I.h If more than one block valve is inoperable, it is necessary to either restore the block valves within the Completion Time of , 1 hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hours ; .d r;;ter; th; The Completion Times are reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation.
- Gr 3 3.i.in l9 2.4.o-+ J G.17 end G.2 and'G:3 If the Required Actions of Condition F are not met, then the plant must be brought to ; "000 in 2.ich the LCO ds;; rct i.,,,1y.
Te-chia; thi; ;t;t;;. the pi;nt .;;t b; bre;.6/. to ai. .aast MOL J within 6 hours and t; "00E 4 T.,, mustJbe reduced: tot <500*F within 12 hours. Additional action;isirequired:;to;be initiated immediately;to'icontinue* efforts to: restore:thetinoperable block valve (s);to 0PERABLE status. This w11U. ensure (expedient seasures areetaken totre-establish-OPERABLE ~ block 1valvestwhile maintaining plantfconditionslabove.MODEE4, butzl_essLthanf500*Fi The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES ? :.r.d 5, 3?(with any_RCS cold tleg ; temperature (368*F),: 4, _ Scand 6:(with the' reactor vessel; head on) mafr.te+ning automatic PORY OPERABILITY may be required. See LC0 3.4.12. SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Blockva/ f1ve cycling _verifie_s that the valve _(s) can be opened and
. io C Dasis for the treauency of 92 cays is the A5ME Code. Section XI (Ref._43)./ the ck y e is los to is ate PORV at
- s bl f bei manu y cy ed. OPE ILI of t bloc y ei of 1 tance aus open g the lock alve es ry to rait _ _P_ORV o be sed fo man con ol of
- on er ve c
- tpex valyens c1 o igflatejwietherwtse pu$per
! (continued) 4 WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 57 5/1/58 7
I l INSERT'B 3.4-57 0 3.4.1.1-4
. The Required Actions are modified 'by a Note stating that the Required Actions
'do not apply if the sole reason for the block valve being declared inoperable is as a' result of power being removed to comply with other Required Actions.
In this event, the Required Actions .for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the Londition. l l e
Pressurizer PORVs
' B 3.4.11 BASES (continued)
SURVEILLANCE SR 3.4.11.1 (continued) - REOUIREMENTS - N g* ,4 ihat.is' pable 7bei ma lly cyc ,t maxi Comp ion T1 o res re t PORV a ope the bl k va e1 urs, ch is I wi in the lowa e limi (25 to i t(7xtend test uir block s wou lve F equency f 92 be c eted ys. F the the r penin of a e, t se r ntly c ed bl valve nr oratio of t PORV
. ERABL tatus -
c letio of the uir Act (fulfi s the SR [( .e. j ]- $ d g h Qates,,p Q S.4.0-2.} N He Note 1 modifies this SR by stating that it is not required to be met performed with the block valve closed in accordance with the Required Actions of this LCO. Note.2: modifies;this rSR[to allow entry linto'and' operation:1 :3 priorato.performingj.,the SR. 7his; allows:.t_heitest"to bei rformed.in MODEE3?under operating temperature and press conditions;-: prior to' entering MODE l'or'2." 4, g g ,w, ps.4.si 4 l con,w. n ie c. renes h A *k -f x umsiclabic. lea k from +He. RG sihca. SR 3.4.11.2 % pen.v 3 sicea.q ihopera ble. . _ _ ~ SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be -- manually actuated for mitigation of an SGTR.fThe Fregnency ofM3A'0" I 8 spffths ifbasepon a typfcal re#0el_10afcyckand4ndustrd aefepted4ractiee./The Note modifies this SR to allow entry into and operation (in MODE ~ 3 prior:to: performing the >SR. This. allows the test to be performed in MODE ~3'under operating' temperature and pressure. conditions, prior to entering MODE 1 or(21In accordance with Reference 5, administrative controls 1requireithis test be performed in-MODE 3 orJ4;to-adequately; simulate operating temperature and pressure effects on PORV operation. - _ O P c"M 8 the rifnca. has shown 4h Ohese. vMvt.s ususth Pgrme$ a+ reqadvd tweeveu.Tesh% Prevem heem,3 y % e , Pre pass ihn.Wsum *"a O* _ -- cz c<wtmb.ta. fi owi o 7sItatr4 t;t:.4 st>nd point. am - 5" 3.;.11.3 Opcratini; the scicacid air centrol valves and check velves on the oir ;;;ni;t;r; cr.;;r;; th; 0'"! ;;r,tr;l ;y;t; ; ;;t;;tc5
. preperly wher, calicd uper. The frcqucacy of [10] r.caths is based on ; typic;l rcfuelin;; cycic and the frequcacy of thc cthcr Survciliences used to d; .eastrate "0'"! 0"E"A"ILI"/.
S" 3.4.11.4 i This Survcilienc; is not rcquired for plants with pccrcacnt IE j p;Wcr ;upplies to the volvcs.
; (continued) 4
( WCGS-Mark-up ofNUREG-1431-Bases 3.4 8 3.4 58 S/VSB7
l CHANGE ! NUMER JUSTIFICATION l l T , measurements are not obtainable for a non operating
- loop.
i 3.4 47 ITS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Condition B or E of LC0 3.4.11. However. Required Action A also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with Nt> REG 1430 and NUREG 1432 in as much as the block valve cycling is exempted under Conditions A.,B. and E.fS e poyer toAbe41ocFValueWu Gsyta J fd ingqui ActMn A.Myhe Note to SR p 3.4 u.+] Qg_ AchD 3.4.11.1 will be revised to not require the survei_llance 4 --
- f 'A. performancece er ifthe the block valve (s) isr closed ock alvps) i 'on ved 'npePjC Re ir Acti s B.2 nd E ,t sur eill ce an t
- t. ive he wo ing ang " met' to " rf rme i N . th wordi of S .4 .1 rev sed o c th[e; at
_he C ition and exc tion.Dhis change is consistent with traveler WOG 87. huerr W9O lQ3 4.H-4] 3.4 48 fA note is added to ITS 3.4.8 ACTIONS indicating that ] entry into MODE 5 Loops Not Filled from H00E 5 Loops h Filled is not permitted while LCO 3.4.8 is not met. The addition of this note is based on the performance of a
' plant specific LCO 3.0.4 matrix (see CN 1 02 LS 1 of QheCTS3/4.0 package). IMan.T c, A . A _
@W$
3.4 49 LC0 3.4.12. "[LTOP] System." provides four differenct methods for pressure relief. Any of the four may be used. However. Surveillance Requirement 3.4.12.5 requires testing whether or not the equipment is being credited to meet the LCO. The proposed change adds the word
" required" to the Surveillance to exempt its performance if the equipment to be tested is not being used to meet the LCO. In addition, two editorial changes were made.
The LC0 requirement presentation was clarified. Also, the Note to SR 3.4.12.8 was revised to replace " required to be met" with " required to be performed" since the l " performed" nomenclature is appropriate here, consistent with the CTS. This change is consistent with traveler l . _
-N (Q 3.411-l l WCGS-Differencesfrom NUREG-1431 - ITS 3.4 9 S/15/97
INSERT 6A-9a 0 3.4.8-1 LC0 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition LC0 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1, 2, 3, and 4. The MODE change restrictions in LC0 3.0.4 were previously applicable in all MODES. ITS LC0 3.4.8 is modified by a Note stating: "While this LC0 is not met, entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is not permitted." The transition from MODE 5 (loops filled) to MODE 5 (loops not filled) removes the steam generators as a decay heat removal system while the RHR System is potentially degraded. Therefore, the Note ensures that the transition is precluded if LCO 3.4.7.b (two SGs) were chosen (in lieu of the second RHR loop) to ensure decay heat removal capability prior to draining the RCS. INSERT 6A-9b 0 3.4.11-4 In addition, Notes are added to Conditions C and F stating that these Required Actions don't apply when the block valve (s) is inoperable solely as a result of its power being removed per Required Actions B.2 or E.3 as a result of an inoperable PORV(s). In this scenario Condition B or E is entered as a result of an inoperable PORV(s). If one PORV were inoperable and incapable of being manually cycled, Condition B would be entered at time zero, to. Required Actions B.1 and B.2 would close the associated block valve and remove its power within time to + 1 hour. If, as a result of block valve power removal per Required Action B.2 Condition C were then entered, Required Action C.1 would require the associated PORV to be placed in manual control within time to + 2 hours. However, the reason for originally entering Condition B is that the associated PORV is inoperable and can't be manually cycled, thus there is nothing to be gained by placing the PORV in manual control. The PORV inoperability may be such that the PORV can't be placed in manual control (e.g., blown control power fuse), in which case Required Actions C.1 and C.2 can't be met. In addition, Required Action C.2 (restore block valve to OPERABLE status) can't be satisfied as long as power is removed from the block valve. Restoring the PORV to OPERABLE status within time to + 72 hours allows the plant to exit Condition B. If power were not restored to the block valve at this time, the new Note on Condition C would have no standing and Condition C would be entered. Similar conclusions can be drawn for the relationship between Conditions E and F. If Condition E is the original Condition entered, there fs nothing to be gained by Required Action F.1 and Required Action F.2 can't be satisfied with block valve power removed. With F.2 not satisfied. Required Action G.2 would require the plant to be in MODE 3, but Required Action E.4 would have already had the plant in MODE 3 two hours earlier.
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 7 of 8 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUtEER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.4-45 ITS3.4 has n vised to ~ ve the e fo Yes No - Operation Yes Yes I[,Requ'edAc on B regardi CCP punp wap ti s of 2 CCPs are l the lic ility No for ac ator i a_t _ allowed per t o ant-s LCO. s discus fic time in trave Iowances r exc ing t LC F) CTS.
-2. \
numbe of [ECCS] s capab of inj ting i o 7 LCSareincor ated. [a_s scussed n CN 3 -18 WE MQ 3.4-46 ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are Yes Yes Yes Yes modified to refer to " operating" RCS loops. . 3.4 47 ITS SR 3.4.11.1 contains a Note which exenpts the Yes Yes Yes Yes cycling of the block valve when it is closed in ^ accordance with Required Actions of Condition 8 or E MAiMn,N 8%es FC 3dded. Ib Omdit$= mF U UN of LCO 3.4.11. However. Required Action A.1 also directs closure of the block valve when one or more
% prevent an Me,1 3u b A p rat 4e PbMQ d
"a" PORVs are inoperable and capable of being manually Y*# I cycled. The SR Note should also exempt performance when the block valve is closed in accordance with l Required Action A.1 as the block valve should not be opened when the PORV is inoperable. =
3.4-48 A note is added to ITS 3.4.8 ACTIONS, indicating that Yes Yes Yes Yes entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is not permitted while LCO 3.4.8 is not met. 3.4-49 This change reorganizes the presentation of ITS LCO Yes Yes Yes Yes 3.4.12, adds the word " required" to ITS SR 3.4.12.5, and changes the word " met" to " performed" in ITS SR 3.4.12.8. WCGS-Conversion Comparison Table-ITS3.4 3/15M7
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.11-5 APPLICABILITY: WC REQUEST: ITS Bases 3.4.11 Background (Wolf Creek) Comment: On the top of smooth Bases Page 3.4-55 the sentence beginning "The functional design..." should not end with " . Pressurizer." It should include the phrase i that ccmprises the next paragraph. l FLOG RESPONSE: The smooth copy of the ITS has been marked to indicate the referenced text as one paragraph. A final review of the smooth ITS and ITS Bases is I planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases. ATTACHED PAGES: None 1
i I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-6 APPLICABILITY: CP, WC, CA , REQUEST: Difference 3.4-49 (Wolf Creek, Comanche Peak and Callaway) l Comment: This difference does not address the addition of the "Immediately"in Required Actions D.1, E.1, and G.1 of ITS 3.4.11 FLOG RESPONSE: This RAI refers to JFD 3.4-49; it should refer to JFD 3.4-39. JFD 3.4-39 is revised with the addition of the following:
"New initial Required Actions are added to Conditions D, E and G to immediately initiate actions for restoration of the inoperable PORV(s)
(and/or PORV block valves) to OPERABLE status. These immediate actions will ensure expedient measures are taken to re-establish the operability of the PORV(s) (and PORV block valves) while maintaining plant conditions above MODE 4 but less than 500 F." ATTACHED PAGES: Encl.6A 7 l I 1 l l l l l
! CHANGE NUMBER JUSTIFICATION purposes (per Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR Noted exception) would not allow the SR to be considered to be met until the appropriate conditions were
- available for it to be performed without entering the l
actions. The Note to these SRs would allow startup in Hode 3 if the SR had not been performed during the required frequency, but would limit the exception to prior to entering Mode 2. The change is consistent with traveler TsTF -@ g z.4.o a_ l 3.4 36 SR 3.4.13.1 and LC0 3.4.15 are revised per traveler TSTF-l 116. The note addresses the concern that an RCS water inventory balan9 e anot be meaningfully performed unless the unit is operc ..,9 at or near steady state conditions. The note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours after re establishing steady state conditions. 3.4 37 Not applicable to WCGS See Conversion Comparison Table (Enclosure 3B). gp 3,g,g _g f_ 3.4 38 __ (~-P' Consis nt hT 105 the d ails o the met l whi t CS f rat are v rified re mov fr ' I.
- hd *PPGm*""
S** '#h N*
.1. to t Bases Movin this i formati t th '
i Bas , al s the se of p cisio at b nc , Q W DW' M th ps, verifi other accepta e met s in o er ion and s cons' tent w' h t N EG 431 p fo
'p ilosop of movi clari ng inf a n pnd ive detat s out the T to Baies.
Qescr , 3.4 39 The shutdown requirements of ITS 3.4.11 would require the plant to reduce T,., to <500 F within 12 hours, rather than MODE 4, to address the concern of entering [LTOP] LC0
- g> 3.4.12 Applicability with inoperable PORVs., For consistency, the shutdown requirements of ITS 3.4.16 are 1414."- 4 also revised to allow 12 hours to reduce T.,, to <500 F.
This change is consistent with TSIF-113.
-QT1F-Z82.]
3.4 40 Consistent with traveler . . ihe Note to SR 3.4.1.4 {Q34.1-2 l would be modified to provi e 3dditional time to perform an RCS precision flow rate measurement. The time allowed c would be changed from 24 hours to 7 days. This change is ( acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and additional time normally would be required to establish WCGS-Differencesfrom NUREG-1431 - ITS 3.4 7 5/158 7
. . - . . . . . . . _. . .-. . . - . .-. . . ..-...-. - . . - . . . ~ . - . - - - . . - - - . - -
l INSERT 6A-7a 0 3.4.11-6 New initial Required Actions are added to Conditions D, E and G to immediately-initiate restoration of the inoperable PORV(s) (and/or PORV block valves) to OPERABLE status. These immediate actions will ensure expedient measures are taken to re-establish the operability of the PORV(s) (and PORV block valves) I' while maintaining plant conditions above MODE 4 but less than 500'F. i
- I l- l l-i I
l l. l + 4 i t l ) t s y gg i--, v , -
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.12-1 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-49 i l Comment: WOG-100 has not yet become a TSTF. I FLOG RESPONSE: WOG-100 has been approved by the TSTF and is designated as TSTF- l 280. This traveler has been submitted to the NRC and is under review. l The proposed wording in TSTF-280 was modified from WOG-100, and these modifications have been incorporated into the ITS (added "or" to
)
i LCO list and SR 3.4.12.5 Note was deleted). The FLOG continues to l pursue the changes proposed by this traveler. I l ATTACHED PAGES: Encl. 5A Traveler Status page, 3.4-29, 3.4-33 Encl. 6A 9 I l l l I 1 . 1 l
- _ . . . - . . . - . . . . - . - -- .. - .- - . -.... - ... - ..~.. .. - . - ~ _ _ . . - _ .. . . . .. - ..
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 TRAVELER # STATUS . DIFFERENCE # COMMENTS l 1 TSTF-26 lacorporated 3.4-32 Approved by NRC. TSTF-27, Revh Incorporated 3.4-33 (Qbyg6((Q 34*L-ll TSTF-28 lacorporated 3.4-22 Approved by NRC. TSTF-54, Rev. I Incorporated NA (App b MMTE..I.+#9l TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not incorporated Minor change that is adequately addressed in the ; Bases. TSTF-87, Rev.h Incorporated 3.4-31 (dp3pd 63 MQ / TIE.3.492V.] TSTF-93 @ Incorporated 3.4-17 {Apprevwd h M /Q.34. iQ) TSTF-94 h Not Incorporated NA Retained current TS. [TM 3.v-6es-l [TS p W 7 /_InmrioratpV f MI4 3 't+ b L TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply. ) TSTF-ll3, RevhI Incorporated 3.4-39 q 3,4,gn 3} TSTF-II4 Incorporated NA Approved by NRC. TSTF-il6, Rev[' Incorporated 3.4-36 l 4.3.4. 85 -4 TSTF-136 Incorporated NA (AP'Fr*'265 NACD /TR 3 #88tl TSTF-137 lacorporated NA Appewysd by MMJ/74 3,4- Sofl TSTF-138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6.. TSTF-151 h Incorporated NA /T4 J.6429] TSTF-153 incorporated 3.4-01 hrwre4 by ME]/723.9-not / TSTF-162 Incorporated NA (Appravv2 by NK.jT3f.3.4- 2*(=} Gl@G4tfBeCD incorporated h3.4-45%5% See also Cns 3.4-18 and 3.4-20.lq 2.4.12.-L T_ (W13C90 TEIFa @ Incorporated 3.4 35 [4 3.4.ll L{ IheC M RamrD incorporated 3.4-10 DCPP onthC;d h AlR3 # o*f I ( WOU-87,'RevM Incorporated 3.4-47 L4 3.4 83-4 } ! M* Incorporated 3.4-40 Applicable to Callaway and ) (Tyrs::-2sg Wolf Creek only. [@ 3.4.1-2.] Incorporated 3.4-49 \Q3.+.l'2.-I$ 5/15M7
LTOP System 3.4.12 i 3.4 REACTOR COOLANT SYSTEM (RCS) i 3.4.12 Low Temperature Overpressure Protection (LTOP) System LC0 3.4.12 An LTOP System shall be OPERABLE with a maximum of cr.c high pic;sur; injection 0ll'D pu;;;p zero safety injection pumps and N M N. one ce' htMfirgal charging ptmp capable of injecting into the RCS and the acctm2iators isolated and cithcr ; or b bclow. one[6f MEE493 thelf_o]TMngIgressureifelieficapabiTities_:
- c. Two "00 cclief vahcs. as fcil;w;.
ta. Two power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, or i Eb. Two residual heat removal (RHR) suction relief valves with -
- setpoints a 436.5 psig and s 463.5 psig, or #N$3ED i
Sc. One PORV with a lift setting within the limits specified f},f in the PTLR and one RHR suction relief valve with a L setpoint 2 436.5 psig and s 463.5 psi Qh4.t2-i E '" 8# 5 i bd. The RCS depressurized and an RCS vent of a 2-07 2;0 square E8?PS M
~~ ~ ' ' "
inches.
...........................N0TES
- - -- g g 2,,, a, y r
hfchhk -h. t
- 1. Twoicentrifugarcharging ptnps may be. capable-of{pdept}en r ipWIM BrIrfet upAo)4' hours for pump ~ swap ~ operation.
b3f4?18) i is.s.4.iz.2 1
~
- 2. Two. safetycinjection pumps _ and two centrifugal charging pinps;may beicapable of~ injecting into the~RCSif(nFin !S3?4s18C W1DElwith;any RCS cold: leg temperature _ s368'f and.ECCS'ptmps
- OPERABLEipursuant.to; LC013.5.2, "ECCSL- Operating,Tand
- (b)
, for upito'4;hoursjaf.ter entering M)DE"4'from M00Er3?or-until the temperature of~o_ne or more RCS coldleg decreases below 325"F,?whichever comes first.
- 3. One or. more safety' injection pumps may be capable (of
^
injecting into theLRCS in-MODES 5 and 6 when the.RCS water '9 314?204 I level ~1s: below the top ofLthe reactor vessel flange.for _ the purpose of' protecting the decay heat 1removarfunction.
- h. _wdsolareA3
- 14. Accumulator htfr if _^a9 rMwhen accumulator 4 374!4514 h pressure 1s8945 er a;=, Gthe maximum RCS pressure for the existing RCS cold leg temperature allowed WM1-2 1 2
by the P/T limit curves provided in the PTLR. WCGS-Mark-up ofNUREG-1431 -iTS 3.4 3.4 29 S/1587
. . - . ~ . . - . - _ _ = . . . _ - . - - - - . . _ . - . - . - - - - - - . . . - - . - - -
LTOP System 3.4.1?. SlRVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY lEjSM38%$ SR M 4;J234 Verify RHR suction isolation; valves fs are open 72;40 hours for each required RER suction relief valve. INI lc? 3.4.il 3 } g l l
~ =. - -
SR 3.4.12.5 -- -- TE - - -- On req ed t pe ormed hen 1
'i q 3 4,ig.q $$NNEs th 3.4 .h d. .
..y ...... ...... ...... .... ..... ... .
l Verify required RCS vent 22.0047 square 12 hours for R31650iG ~ l inches open. unic;ked egn l vent velv;';) l pathway (sHnat l lockedCseaT6d or;ot.herwise secured inithe l l openl position I l \
- M 31 days for 1;;ted ;g , %314
- 508 vent valve (s) i locked /1 sealed i or otherwise l secured i_n the l open position 1
SR 3.4.12.6 Verify PORV block valve is open for each 72 hours required PORV. l l l SR 3.4.12.7 Not Used V;ri'y ;;;;;isted IJir, ;uction 31 day; gyp @rd i;;1; tion volv; i; 1;;k;d egr. .;ith egr;ter g'.;;r i; ;.;d 'er ;;;h r;quir;d IJ:I, ;uctier. E-D __,2.s_,.
.s. .... .
.g 3,4 ,L,3 1
WCGS-Mark-sp ofNUREG-1431 -ITS 3.4 3.4 33 S/15M7 1
i CHANGE NUIEER JUSTIFICATION T, measurements are not obtainable for a non operating loop. ! 3.4-47 ITS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with ; Required Actions of Condition B or E of LCO 3.4.11. However Required Action A also directs closure of the block valve when one or more PORVs are inoperable and j capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in ! accordance with Required Action A.1 as the block valve ; should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 in as much as the b'ock valve cycling is exempted under _ tcuthealocFValueWu Conditions A._B. and E.TS ei Gs me1Titginfd irj)Nfqui Ac nA.XJfheNotetoSR 3.4.11.1 will be revised to not require the surveillance @ 3.4u.+] Qg_ Q kM u(. o e i LCO . j performance if the block valve (s) is closed Def)C 1 on
%. ce r the ock alv s) r ved n i R ir Acti s B.2 nd E ,t sur eill e an t
- t. ive he wo ing a " met to " rf 1 t
,t wordi of .4 .1 re sed o e e q ition exc tion.Dhis change is _
C and consistent with traveler WOG 87. auseAT 4,4% le 3.4.u-4j 3.4 48 IA5te 1's added to ITS 3.4.8 ACfiuNS indicating that] entry into MODE 5 Loops Not Filled from H0DE 5 Loops % Filled is not peMitted while LCO 3.4.8 is not met. The addition of this note is based on the performance of a l
? plant specific LCO 3.0.4 matrix (see CN 102 LS 1 of i D CTS 3/4.0 package). luma.T c,A -94 _ .
@W$
3.4 49 LC0 3.4.12. "[LTOP] System." provides four differenct methods for pressure relief. Any of the four may be used. However. Surveillance Requirement 3.4.12.5 requires I testing whether or not the equipment is being credited to l meet the LCO. The proposed change adds the word
~
" required" to the Surveillance to exempt its performance if the equipment to be tested is not being used to meet the LCO. In addition. two editorial changes were made.
The LCO requirement presentation was clarified. Also, the Note to SR 3.4.12.8 was revised to replace " required to be met" with " required to be performed" since the
" performed" nomenclature is appropriate here, consistent i
wit _h the CTS. This change is consistent with travele.r N- h ) LQ 3 A.11-1 r WCGS-DifferencesfromNUREG-1431-ITS3.4 9 $/15/97
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.12-2 APPLICABILITY: DC, CP, WC, CA l REQUEST: Di*ferences 3.4-23 and 3.4-45 Comment: WOG-51 Rev.1 has not yet become a TSTF. ! FLOG RESPONSE: WOG-51, Rev. 2 has been approved by the TSTF and is designated as TSTF-285. This traveler has been submitted to the NRC and is under i review. The proposed wording in TSTF-285 was modified from WOG-51, Rev. 2, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Encl. 2 4-34,4-34 Insert page I Encl. 5A Traveler Status page,3.4-29 , Encl. 58 B 3.4-67, B 3.4-68 j Encl.6A 8 l Encl. 6B 7, 8 l l l l l 1 l 1 1 l l
J l OVERPRESSURE PROTECTION SYSTEMS - LIMITING CONDITION FOR OPERATION
, 3.4.9.3 At least one of the following groups of tw6 overpressure protection r
devices shall be OPERABLErwith a maximum of zero safety injection pumps 1 9-06-M i- and one centrifugal charging pump capable of injecting into the RCS and the accumulators isolated or depressurized below allowed RCS pressure per the j 9 PTLR Iwhen the Reactor Coolant System (RCS) is not depressunzed through a 3
- 2 square inch or larger vent:
. a. Two residual heat removal (RHR) sucLon relief valves with Setpoints of 45013%, or
- b. Two power-operated relief valves (PORV) with Setpoints which do not y
exceed the limits established in[the PTLR,7!;rre 2.' ', or 9-01-LG
- c. One RHR suction relief valve and one PORV with Setpoints as prescribed g
%3.4 ll-L}
APPLICABILITY: MODE 3 when the temperature of any RCS cold leg is less than
- or equal to 368'F, MODE 4, MODE 5, and MODE 6 when the head is on the Reactor Vessel.
L f INSERT 3.4.9.1 1} 9-17-LS-24 , ACTION: i
- a. With one of the two required overpressure protection devices
- inoperable in MODE 3 or 4, restore two overpressure protection devices to OPERABLE status within 7 days or depressurize and vent the RCS j through at least a 2 square inch vent within the next 8 hours.
l b. With one of the two required overpressure protection davice4 , inoperable in MODES 5 or 6, restore two overpressure protection E ! devices to OPERABLE status within 24 hours, or complete ! depressurization and venting of the RCS through at least a 2 square i inch vent within the next 8 hours. 1
- c. With both of the two required overpressure protection devices inoperable, complete depressurization and venting of the RCS through i at least a 2 square inch vent within 8 hours.
i: 1 '- 'h: x=t :2:: 'h: POP'/:, :: 'h: P" -~"': . ri' r:! =, er 9-07-TR-2 '
'5: P.CS =';:) ::: tr ' t: .M;::: = "CS ; - r : i::rt!=', c
- ^ :'.-" h ;. : ^^d"I-'^^ I'.0 $2 C - :^J .
S;:il' ":;?;:f-p=- _m ' a- :, e.0.2 ",20 d:y:. The ::;: 1 :S:!!
:::t "h: :!:r 12 :z :^^'.; 'h: t- ri ', 'h: :"xt f '50 3
"OP'!:, e:'h: P"" rd ri' :frx, :: .CS =t;:) : . 'he
.:;--- r ', x ': / x: ntz :" :- :::- : / t: rr' r:r :xx.
i . e. The provisions of 3 0.4 are not a- .R*:=
- j. (NEW) With one or more safety injection pumps or more than one centrifugal ' 9-15-M
, charging pump capable of injecting into the RCS, immediately initiate 1 action to verify a taaximum of zero safety injection pumps and a maximum j of one centrifugal charging pump is capable of injecting into the RCS or , depressurize and vent the RCS with an RCS vent of 22.0 square inched 1 ( within the next 8 hours. j (NEW) With an accumulator not leolated when the accumulator pressure is greater 9-10-M j than or equal to the maximum RCS pressure for the existing cold leg
- temperature allowed in the PTLR, isolate the affected accumulator within 1 hour or, within the next 12 hours, either increase all RCS cold leg temperature j- to >388'F or depressurize the affected accumulator to less than the maximum k [ RCS pressure for existing cold leg temperature allwed in the PTLR. J F(NEW) With the LTOP inoperable for any other reason, within 8 hours depressurize l 9-11-M the RCS and establish RCS vont of 22.0 square inches.
WOLF CREEK- UNIT 1 3/4 4-34 Amendment No. 63
- Mark-up ofCTS 3M.4 SMSM7 i.
(INSERT 3.4.9.31] made Q 3.4.12. -2. } I NOTES
- 1. Two centrifuaal charging pumps may be capable of injecting @ 9-17-LS-24 SeitCJep444 hours for pump swap operations.
,{ C4*r 4)
- 2. Two safety injection pumps and two centrifugal charging pumps may be capable of injecting into the RCS:
(a)in MODE 3 with any RCS cold leg temperature < 368* F and ECCS pumps OPERABLE pursuant to LCO 3.5.2, "ECCS-Operating", and (b) For up to 4 hours after entering MODE 4 from MODE 3 or I until the temperature of one or more RCS cold legs decreases I below 325* F, whichever comes first.
- 3. One or more safety injection pumps may be capable of injecting into the RCS in MODES 5 and 6 when the RCS water level is below the top of the reactor vessel flange for the purpose _of protecting the decay heat removal
'a"cuaa-goe.muount) (3 iw ;
- 4. Accumulator ^ ; -aZ..., . _ _7when accumulator pressure is Q 3.4.L'4 - 2. } I (tharrqGequaPts)the temperature allowed bymaximum the P/T limitRCS curvespressure provided infor thethe existing RCS cold leg PTLR. i I
1 Mark-up ofCTS 3M.4 5/15/97
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-26 incorporated 3.4-32 Approved by NRC. TSTF-27, Revh incorporated 3.4-33 (Ag.3b 14QG14."L-il 3 TSTF 28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev.1 Incorporated NA (AFP k *J M ne.5.&m91 TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not Incorporated Minor change that is adequately addressed in the Bases. TSTF-87, Rev.h incorporated 3.4-31 (dplpe 63 hQ /rlt3.4,24. j TSTF-93 @ Incorporated 3.4-17 h yre d 6h hit Q /9 3.4. i3] _ TSTF-94 h Not Incorporated NA Retained current TS. [TM 3.4-m s-l (TS p V 2 21n d ratKf M' ! 4 3.4. l' 1 TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply. TSTF-ll3, RevhI Incorporated 3.4-39 q 3.4gg 3] TSTF-ild Incorporated NA Approved by NRC. TSTF-il6, Rev[' Incorporated 3.4-36 l 43.4. 83-3_.] TSTF 136 Incorporated NA (APPrwuL 63 NACD /72.3.680t f TSTF-137 incorporated NA @ppe ved 63 AAEJ/7A 3.#- Sof l TSTF-138 Not incorporated NA Inconsistent with RCS loops : requirements ofITS 3.4.5 and 3.4.6.. TSTF-151 Q incorporated NA /T4 J.#4pfl TSTF-153 Incorporated 3.4-01 hM by AMQ/723.Pnoy / TSTF-162 incorporated NA (AppArveJ gg Tit. J. 4- sor.] GEOGM:'Ref"D See also Cns 3.4-18 and 3.4-20.19 84il~ M Incorporated h3.4-45%s)2. (WIVp90 'TilF*@ Incorporated 3.4-35 [4 3.4.tl .'L.1 Bhf8C49:"Ramro incorporated 3.4-10 DCPP onid;f. .S by @M3.Mt I ([WOG-87,'Rev3 Incorporated 3.4-47 [ 4 3 . 4.83- 4 ) M* Incorporated 3.4-40 Applicable to Callaway and (T1rrs:-28@ Wolf Creek only. [@ 3 4.1-2.] Incorporated 3.4-49 L 4 3.+.I'3. - Il S/158 7
LTOP System 3.4.12 l 3.4 REACTOR C00UWT SYSTEM (RCS) l 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of are l;ip, prs =r; ir.jati; 'l'."!) p;;;;p zero safety injection pumps and NON one 6^.iT aii&] s charging pump capable of injecting into the RCS and the accumulators isolated and ;itter ; er b b;ls. onefof E$$@ LIM theffoWaiHngZpressureirelieficapabilitiesY l :. Tw ",CS reli;f v;lva. ;; f;11ri ;. l -la . Two power operated relief valves (PORVs) with lift settings l within the limits specified in the PTLR or ab. Two residual heat removal (RHR) suction relief valves with setpoints a 436.5 psig and s 463'.5 psig, or %M3 Sc. One PORV with a lift setting within the limits specified [},g l in the PTLR and one RHR suction relief valve with a L
%3.4.t2-T W8dd setpoint a 436.5 psig and s 463.'5 psi bd. The RCS depressurized and an RCS vent of 2 0-07 2;0 square gj;g; ppg inches. A p z.4.a- 2.i h, u_d _*ng b .h.
...........................N0TES - ------ -- -
gg 1~. Twojcentrifugarcharging. pumps may be capable. of(odept1sn t r Trikf 26 BGrferAAo)4 hours' forgptsup swap operation.
- ~
g3ygg p .s.4. a.-M 2
. Two;safetyfinjection pumps and.two. centrifugal; charging _ . . . .
~
l pumps may belcapable of~ injecting;into the1RCSi;(alin 9 314318 % l MODE;3 withiany RCS~ cold1 1eg temperature s368'F;and ECCS' pumps j OPERABLE! pursuant,to LCO 3.5.2EECCS1- Operating."Tand (b) ( for'up"toi4'hoursjafter enteringLMODE'4ifrom NODEi3.orj_until l the temperature of one;or more_.RCS~ cold leg decreases below 325'FT whichever comes first.
- 3. One or more safety injection pumps may be_capablef of ..
l injecting;into the RCS in MODES 5 Land 6 when the RCS water F3 W 204 level is;below the top of the. reactor vessel. flange.for the' purpose of protecting the decay heat removarfunction. Aho\>&J
- 34. AccumulatorbMM M aaNDwhen accumulator 5314M534 h pressure ish:5sFt:._~. r 0;=: ~;;2,the maximum RCS pressure for the existing RCS cold leg temperature allowed WM-2 1
' ' ~
. by the P/T limit curves provided in the PTLR. l f l WCGS-Mark-up ofNUREG.1431 -ITS3.4 3.4 29 S/15/97 1 I
I LTOP System B 3.4.12 BASES APPLICABLE RCS Vent Performance SAFETY ANALYSIS (continued) With the RCS depressurized, analyses show a vent size of E d7 l 2!D square inches is capable of mitigating the :11;nd LTC" limiting >LTOPij cacp ;m.; transient. The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, two oneicentrifugal~ charging pumps OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve. { The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the i vessel material surveillance. The RCS vent is passive and is not subject to active failure. l The LTOP System satisfies Criterion 2 of th N"C l'clicy l l Statement.-TIO CFR250~367(c)(2)(11) . . ! l LCO This LCO requires that the LTOP System is OPERABLE. The LTOP l System is OPERABLE when the mWate maxima coolant input ors heat i input bounded.by<thatiassmed in the. analyses and required I pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient. l To limit the coolant input capability, the LC0 requires that a l maximum of zero safety injection pumps and two one centrifugal @Mn.-2) l charging pumps be capable of injecting into the RCp'nd all ac lator discharge isolation valves be closed aht! immobilized
- accumulator pressure is greater than or equal to the
[ max mum RCS pressu for the existing RCS cold leg temperature
- allowed in the P I
l The LCO _is; modified by four Notes. Note _1 allows.two centrifugal L charging pumps,tofbe; capable of: injecting:1_nto.the RCSifor <14
~
- hours for
- pimp swap operations. This:provides thennecess_a_ ry_ .
l allowance t7 oiperform the ptmp swap activities @MLed@rtDhid_.ps.4.n.-2. l - pagnerland provides: sufficient. time to1 complete thetactivities necessary to^ restore a<maxista of one c.entrifugal charging:ptap j fte'a4$stad)capableLof l injectingM This .1s accomplished by. racking out the breaker for one pump;orreuploying f j (continued) WCGS-Mark-sqp ofNUREG-1431-Bases 3.4 8 3.4 67 $/1/SM7
I l LTOP System l B 3.4.12 l 1 BASES l LCO two; independent means to prevent a pump start in accordance with
'(continued) SR'3:4'.'12:2 Note 2 recognizes the Applicabi_11ty._.ov_erlap.:be_ tween LCO's,3.4'12 .
and;3.5:2^and_. states that; two safe _ty injectioniptap_slanditwo centrifu~ gal charging pumps may be capable;of= injecting intolthe RCS{ gg gg,g (a) In MODE 3 with any RCS cold leg temperature < 368' F and ECCS pumps OPERABLE pursuant to LC0 3.5.2, "ECCS-Operating", and (b) For _up ,to 4 hours after entering MODE 4;from MODEJ or_the temperature;of~one,or more RCS1coldflegsZdecreases.'below 3257,' whichever comes first. Not_ei31statesithat oneLor more_ safety injection _ptaps;may bec_apable_ offinjecting intojhe_RCS in MODES _5_and;6 when the RCS 1 6LM~Ithje jjnp,M@ig P M se 3fT tik: ting ~theldecaylheat' removar functioni [~C"' % *? "'" ?*' ***D p3Aa.2.\ Note 14jstates_that;acctmulator s when the: accumulator pressure 13, the maxistm g=, 4m ___ - RCS pressure _for the ex1. sting RCS cold.]egit~emperaturelas' allowed bytthe~PO1mitfcurves provided in the~PTLRT~This: Note': permits theiaccumulator, discharge isolation valve Surveillance to. be Perfpraedionlyfunder these pressureland1 temperature; conditions. The elements of the LC0 that provide low temperature overpressure mitigation through pressure relief are:
;. Two RCS rclief volyc;. c; follow;.
la. Two OPERABLE PORVs: or A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint. l and motive power is available to the two valves and their (D *MM*^O control circuits. reAhr.u 2b. wo OPERABLE RHR suction relief valve 1 (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 68 S/1/S/97
! CHANGE NUMBER JUSTIFICATION plant conditions suitable for the precision heat balance. l Since this parameter does not normally change l significantly and the flow meters can be used in the l interim, there is no need to perform this SR within the 24 l hour period specified in NUREG 1431 Rev.1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition, the THERMAL POWER specified in the Note would be changed from the gei.eric value in brackets (90 % RTP) to 95 % RTP. This change is acceptable because it specifies a power level in better agreement with current operating procedures for performing a precision l heat balance. Current TS do not specify a power level for this measurement. l 3.4 41 Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 6B). 3.4 42 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). 3.4 43 A new Condition is added to LCO 3.4.1 to reflect the current licensing basis of Wolf Creek for RCS flow rate. r License Amendment 61 approved revisions to incorporate the provisions of the RCS flow TS entitled "RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR" into the "0NB PARAMETERS" specification. These changes were made to support the use of VANTAGE 5H fuel with the Intermediate Flow Mixer grid feature. This amendment also approved L operation at an increased power level. M do x.4G 2 4 s-2 3.4 44 N o WCGS -1iee f.omr6rsiop-Gol!Ipar-tsOfi Tabie] cl re . mWGM lG 3.4.12,2._ f 3.4 45 i ITS 3. 2 has" een re sed move he Not for equ ed I l Acti B.1 r ardin CP p p swap operat ns dt ' l . licabi ty Not for a umulat isol ion th LCO. as dis ssed 1 ravel WOG 5 '. Rev. . P1 nt- cif 4 l j ti 11owan s for xceedin the LC *s n er f [E S] ! s cap e of jectin RCS e i corpo at , i as disc sed in N 3.4- . ntot[Not Th 'e de il s) ua ons wher except ns to t LC0 ar permi ted nd ap6 e j a ropriat y annot ed und the L _ lOSEd.T GA-8a ) , 3.4 46 Consistent with current TS 3/4.1.1.4. " Minimum Temperature - for Criticality," ITS LC0 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. i- Adopting the current TS wording is acceptable since valid S/lSi97 WCGS-Differencesfrom NUREG-1411 - ITS 3.4 8 l-1
- ._. = . - _ . ~ .. - . - . .
INSERT 6A 8a 0 3.4.12-2 Consistent with traveler TSTF-285, ITS 3.4.12 has been revised to move the Note for Required Action B.1 regarding CCP pump swap operations and the Note under Applicability regarding accumulator isolation to the LCO. These Notes have been reworded for clarity and detail situations where exceptions to the LC0 are permitted. Also, plant-specific time allowances for exceeding the LCO number of [ECCS] pumps capable of injecting into the RCS are incorporated [, as discussed in CN 3.4-18). INSERT 6A-8b 0 3.4.5-2 0 3.4.5-3 Steam generator levels for MODE 3, 4, and 5 are specified to ensure SG tubes are covered. The current TS did not ensure tube coverage. l l I 1 i t t l
i
' CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 7.of 8 J SECTION 3.4 L i
DIFFERENCE FROM NUREG-1431 APPLICABILITY I NUNBER DESCRIPTION DIABLO CANYON CONANCHE PEAK WOLF CREEK CALLAWAY : , 3.4-45 (ITS 3.4 ~~ has " vtsed to ~ ve uiii e Yes No - Operation Yes Yes
,Requ on regardi CCP pump ti of 2 CCPs are j the lic 111ty No for ac ator i lati -
allowed per !
>o LCO. discus in trave % / p ]s* CTS.
. @ -2. _j i
nt- fic time 10wances exc ing t L - i of [ECCS] s capab of in ting i o 7 i
@ are incor ated. [as scuss n CN 3 1 spE M-{
3.4-46 ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are Yes Yes Yes .Yes modified to refer to " operating
- RCS loops. ,
j 3.4-47 ITS SR 3.4.11.1 contains a Note which exempts the Yes Yes Yes Yes cycling of the block valve when it is closed in - . accordance with Required Actions of Condition B or E b ddihAn,N 8%e6 Tre. added.
- 0mattia. C. mF M #4 of LCO 3.4.11. However. Required Action A.1 also 9 ,pfevent an 31 Aae b o g otAm PbAvQ directs closure of the block valve when one or more I PORVs are inoperable and capable of being manually D*# D *" a # !
cycled. The SR Note should also exempt performance i t when the block valve is closed in accordance with Required Action A.1 as the block valve should not be i I opened when the PORV is inoperable. w- ! 3.4-48 A note is added to ITS 3.4.8 ACTIONS. indicating that Yes Yes Yes Yes i entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is not permitted while LCO 3.4.8 is not met. t 3.4 49 This change reorganizes the presentation of ITS LCO Yes Yes Yes Yes 3.4.12. adds the word " required" to ITS SR 3.4.12.5. i and changes the word " met" to " performed" in ITS SR 3.4.12.8. WCGS-Conversion Congparison Table-ITS3.4 S/15/97 i ___._.._m__.. - _ _ _ _ _ _ _ . . _ _ . _ . - .__ . _ _ . _ _ _ :umm._-._______..._.i
INSERT 6B-7a 0 3.4.12-2 TECH SPEC CHANGE APPLICABILITY DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER 3.4-45 Consistent with traveler TSTF-285, ITS 3.4.12 has been revised to move the Note for Required Action B.1 regarding CCP pump swap operations and the Note under Applicability regarding accumulator isolation to the LCO. Also, plant-specific time allowances for exceeding the LCO number of [ECCS] pumps capable of injecting into the RCS are incorporated [. as discussed in CN 3.4-18]. I s l A \ L
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 8 of 8 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY t 3.4-50 This change is consistent with current TS SR No - adopting No - adopting Yes Yes 4.4.9.3.3. The 12 hour frequency applies to vent ITS format. ITS format. pathways that are not locked, sealed. or otherwise secured in the open position. The wording added to ITS SR 3.4.12.5 is also consistent with the format ; used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. 3.4 51 The Note for SR 3.4.1.4 is removed. This is Yes No No No consistent with DCPP CTS 4.2.3.5. DCPP conducts measured RCS total flow rate verification on the month frequency. lDC ALl.-005") L 3.4 52 Consistent with traveler the Note concerning No - See CN Yes No - See CN No - See CN h3.4.nl-2. ] accumulator isolation is ved from the APPLICABILITY 3.4-45. 3.4-45. 3.4-45. to the LCO. [- -28D _ m_ --
-Creierdea forct2r%~ M h)
(34-s3 mseAT c,a uc.3.4-cor} (g.4_54- iwse2.T 6B- 8&[ oc. 3.it-cos ] WCGS-Conversion Comparison Table-ITS3.4 5/15/97
INSERI 68-8a WC 3.4-007 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 3.4-53 This change revises the ITS SR 3.4.11.2 No No Yes Yes Frequency from "18 months" to "In accordance with the Inservice Testing Program." The CTS for this surveillance establishes the frequency as being per the IST Program [ CTS SR 4.4.4.1]. INSERI 6B-8b DC 3.4-003 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY No 3.4-54 Consistent with the current license bases Yes No No as approved LA 124/122. ITS LC0 3.4.13 is revised to reflect reduced steam generator primary-to-secondary leakage limits of 150 gallons per day from any one steam generator and an additional surveillance Requirement to determine primary-to-secondary leakage every 72 hours.
l ADDITIONAL INFORMATION COVER SHEET 4 ADDITIONAL INFORMATION NO: O 3.4.12-3 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-09 ' Comment: The difference does not adequately justify not adopting STS SR 3.4.12.7. The SR is intended to apply to valves besides manual valves. Performing SR 3.4.12.4 does not verify the same status as that verified by SR 3.4.12.7. j l FLOG RESPONSE: JFD 3.4-09 is not applicable to DCPP. l JFD 3.4-09 provides an incorrect justification for not adopting SR 3.4.12.7. The Surveillance Requirement to verify the RHR suction isolation valves locked open every 31 days (when the RHR relief valves are being used for overpressure protection) was removed from the CTS as part of a ; license amendment implementing the Generic Letter 88-17 recommendation to delete the RHR autoclosure interlock (ACl). The 31 l day surveillance was determined to be no longer necessary since removal of the ACI eliminates the single failure that could have isolated both RHR suction relief valves. ACI removal also reduces the probability of closure l of the RHR suction isolation valves when power is available. Also, SR 3.4.12.7 is bracketed in the STS. NUMARC 93-03, " Writer's ! Guide for the Restructured Technical Specifications" indicates that brackets are used in the generic Technical Specifications and Bases to ; indicate where plant specific input is needed. As identified in the i , " Methodology for Markup of NUREG-1431 Specifications"in Enclosure SA, changes to bracketed inforrnation involve the insertion of plant specif5::information which is presently located in the current TS. The methodology applied by the FLOG was that a JFD was not required if the ; bracketed requirement /information was not in the current TS. Therefore, i no justification was provided since the STS SR 3.4.12.7 was not in current l TS. SR 3.4.12.4 is also bracketed in the STS. The changes being made to that surveillance involve plant specific wording changes (i.e., " isolation valves are"), which require no justification per the FLOG methodology, and the SR Frequency as discussed in JFD 3.4-08 (not applicable to DCPP). Based on the above, JFD 3.4-09 is no longer necessary and will be replaced by "B-PS' in the Enclosure SA markup. JFD 3.4-09 will be rhown as "not usd in the Enclosure 6A and 6B markups. Plant Specific Discussion: ACI deletion and elimination of the subject surveillance requirement was approved for Wolf Creek in Amendment No. 49 dated September 12, 1991. i
l ATTACHED PAGES: Encl. 5A 3.4-33 Encl. 6A 2 Encl.6B 2 l
LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS (continued) , SURVEILLANCE FREQUENCY ERf3M30810 SRC374:12:4
~
~ ~ ~ ~ Verify RHR suction isol.ation[ valves +s arejopen 7E4B hours for each required RHR suction relief valve. IM. _ )
lQ 3.4.it.3 / SR 3.4.12.5 -.. ... .... . .. .... .. 1th 3.4 .b d. .
^
lM: _ BEPSM$ Verify required RCS vent 22.0 E-97 square 12 hours for 4374150Mi
~~
inches open. ur. led d agr. vent valvc(;) pathway (s)fnot locked.Tse8Ted orj otherwise securedfinithe open position htD 31 days for 1;d:d ;gr, s#3f4!50% vent valve (s) locked, sealed or otherwise secured in:the open position SR 3.4.12.6 Verify PORV block valve is open for each 72 hours required PORV. SR 3.4.12.7 Not Used Verify ;;;;;icted rJir, sucticr. 31 day; ty g Q i;;isticr velva i; ledcd cgr. with cgretor ia.;;r iceved for ced rcquired PJ:P, suction B" >[ Miicf volvc. hG 3,4 ,,,3 WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 33 S/1S/97
a l 2 CHANGE l Nl#EER JUSTIFICATION
; 3.4 05 Not applicable to WCGS. See Conversion Comparison Table i (Enclosure 68).
6 : i 3.4 06 These changes are consistent with the current plant specific analysis and are reflective of the current ! Technical Specifications. The plant specific analysis does not allow injection from any safety injection pumps and does allow injection from [one) centrifugal charging pump. The changes are made consistently in the LCO, ACTIONS and SURVEILLANCE REQUIREMENTS. ! 3.4 07_ Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 3.4 08 The existing licensing basis as contained in the Technical l Specifications requires performance of this surveillance on a frequency of 72 hours. The Westinghouse STS used to develop the plant specific TS did not address the use of ' RHR relief valves. The requirement in the current TS was developed as part of an LAR to remove the autoclosure interlock which, in part, proposed 72 hours as it was consistent with the SR for the pressurizer PORV block valves. The 72 hours was found to be acceptable in the SER which was enclosed in the license amendment. Plant j experience has not indicated that the existing requirement 1 is unsafe or unacceptable. The surveillance frequency does not require reduction to 12 hours. __ _ J 4 3.4.12-3 3.4 09 k _ nt dop (not e ma 1 RHR uctio isol on ! va es. TM moto opera d suct n iso tion alve (2 r rel f val line) re sur ill n ac rdan wi t i SR 3. 12.4 Theref e this surve ance equi n isj ' ed. t4at use:L.y 3.4 10 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). , 3.4 11 The plant does not have the RHR autoclosure portion of the RHR System interlock as the system was deleted from the plant design. However, the portion of the interlock which prevents the valves from opening when system pressure d in excess of the setpoint has been retained. As such the i note referring to the autoclosure interlock has been i deleted from improved TS 3.4.14 Condition C and SR 3.4.14.2 and SR 3.4.14.2 is modified consistent with LC0 3.4.12.. SR 3.4.14.3 is not used. WCGS-DifferencesfromNUREG-1431-ITS3.4 2 S/13/97
m.
,.-m. .in.
CONVERSION CdMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 2 of 8 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUPEER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.4 08 The current licensing basis as contained in the Nu - DCPP LTOP Yes Yes - See Yes - See OL Technical Specifications requires performance of this design does not Amendment No. Amendnent No. surveillance on a frequency of 72 hours. use RfR relief 49. 42. valves. 3.4 09 r "The pl does t hav nua s tion 01at' No[DCP L yee *JA yes' MA Joe" % i val- . The tor- rated ucti isol ion v ves, A si does t ( per rel f valv line re s eill in a cordshce use r ief >" '~ - u ith SR _.4.12. 64at used. _ v ves. >J Af g-- - LjQ 3.+.82 -3 ] 3.4-10 The DCPP plant specific limiting temperature specified Yes No No No in degrees below which the RCS must not be subject to low temperature overpressure is replaced by the generic statement "the temperature below which LTOP is required as specified in the PTLR." 3.4-11 The plant does not have the RfR autoclosure portion of No - The valve Yes Yes - See Yes - See OL the RfR System interlock as the system was deleted interlock is Amendment No. Amendment No. from the design. However, the portion of the not in the 49. 42. interlock which prevents the valves from opening when current TS system pressure is in excess of the setpoint has been retained. 1 p. 3.4-12 In conformance with the current TS. the RIR Isolation No - RIR vah e , (e ) No - WCNOC No - Callaway Valves which are RCS PIVs are excluded from being testing after i dees not have does not have retested following an extended period of operation in MODE 5 8 this this H0DE 5. operation is exclusion. exclaston. not in CTS. 3.4-13 The PJR Isolation Valves which are RCS PIVs are No - This Yes Yes Yes excluded from being retested following flow through change is out the valves. of scope for DCPP. WCGS-Conversion Comparison Table-ITS3.4 S/lS/97 l _-____ a
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.12-4 APPLICABILITY: WC, CA, CP REQUEST: ITS Bases 3.4.12 Applicability (Comanche Peak, Wolf Creek, and Callaway) Comment: The intent of the addition to the end of the first paragraph of the Applicability Bases is unclear. The LCO applies if the head is on. The added discussion essentially states LTOP (COMS) protection is not needed with the head on and the bolts fully detensioned. If that is the argument then rather than adding it to the Bases discussion, the case should be made for modifying the LCO Applicability. FLOG RESPONSE: This comment is not applicable to Comanche Peak as this information was not in the CPSES ITS Bases. The Applicability for ITS LCO 3.4.12 includes MODE 6 when the reactor vessel head is on. With no fuel in the reactor vessel, the plant is not in MODE 6. The statement was placed in the ITS Bases to indicate that low temperature overpressure protection (LCO 3.4.12) is not required to be OPERABLE with no fuel in the reactor vessel. There may be some plant conditions when the reactor is defueled that warrant placing the reactor vessel head on the vessel for radiological concerns. In these situations, the requirements of LCO 3.4.12 are not required to be met. The inserted Bases words are being deleted and , plant procedures will provide the appropriate guidance for the plant conditions when no fuel is in the reactor vessel. ATTACHED PAGES: Encl. 5B B 3.4-69
l 1 l LTOP System l B 3.4.12 I BASES LC0 "An RHR suction relief valve is OPERABLE for LTOP when ith (continued) RHR suction isolation valves and it; "J:", ;;; tion volv; are - open, its setpoint is at or between 436.5 psig and 463.5 psig, and testing has proven its ability to open at this setpoint. . 3c. One OPERABLE PORV and one OPERABLE RHR suction relief give;or j {QM am-l } re.dhnt. 1 bd. A depressurized RCS and an RCS vent. (em McA % I An RCS vent is OPERABLE when open with an area of :t 0-07 2.0 square inches. l Each of these methods of overpressure prevention is capable of l mitigating the limiting LTOP transient. l APPLICABILITY This LCO is applicable in H00E'31when the tyr;tur; cf anylCS coldJegitamperature~isi 368"F, in H00E 4..;h;n ;ny "CS ;;1d 1;i; tWrotur; i; 4275]"i in H00E 5 and in H00E 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits in H0 DES T.12,Tand3 ab;;; 275'r. When the reactor vessel head is_ j off,overpressurizationcannotoccurjWit fuel f1 ded, he\ rea_cto essel ad_ . be. _ aced the; _self or2r _iol c f co _ ions. _ not 7ted Ove essure rot ifon_ ' ma a l i __use n .rces or the 11y uced _erpre _ure _rei i avail eTand rea or v 1~hea will: ft re ev ' at ( risiu 'i f 2 pre ufe ~T
~
1f T ~auTica p L lq s.4.12-+ } LCO 3.4.3 provides the operational P/T limits for all H0 DES. LC0 3.4.10. " Pressurizer Safety Valves " requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2 and 3. and "00: 4 abov; 320*I. Low twperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event. l l l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-69 S/1/587
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.12-5 APPLICABILITY: WC, CA REQUEST: Differences 3.4-18 and 3.4-45 (Wolf Creek and Callaway) Comment: The justification for the 4-hour pump swap is inadequate. The STS allows 15 minutes. The CTS is used as justification however, finding a pump inoperable and then restoring it (which is the case covered by the CTS) is very different than simply j switching from one operable pump to another. ' FLOG RESPONSE: Four hours is a reasonable time restriction for swapping centrifugal charging pumps (CCP) during the low temperature overpressure protection (LTOP)/ cold overpressure mitigation system (COMS) Applicabiiity. Current Technical Specification (CTS) 3/4.4.9 Bases state " Operation below 350F but omater than 325F with all centrifugal charging and Safety i Injection pump; 9ERABLE is allowed for up to 4 hours. . Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed. Initiation of both trains of Safety , injection during this 4-hour time frame due to operator error or a single ! failure occurring during testing of a redundant channel are not considered to be credible accidents." Additionally, CTS 3.5.4 requires all Safety injection pumps and one CCP to be inoperable. If this requirement is not met, then four hours is allowed to return the pump (s) to an inoperable status. Performing CCP swap operations for maintenance activities requires both pumps to be capable of injecting for a limited period of time. During the : time allowed for pump swap operation, the inoperable CCP must first be restored to OPERABLE status to meet ITS LCO 3.5.3 (MODE 4) and USAR/FSAR Section 16.1.2.3 (one OPERABLE CCP in boration flow path, MODES 4-6). Then the other CCP must be rendered capable of injecting. In order to render the other CCP incapable of injecting into the RCS, the requirements of ITS SR 3.4.12.2 must be met. SR 3.4.12.2 Bases states that a pump is rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out l under administrative controls. The Bases also state that an alternate method of cold overpressure protection control may be employed using at least two independent means to render a pump incapable of injecting. Each method includes local actions (e.g., breaker racked out and tagged. valve closed and tagged). These actions for restoring the one CCP and then rendering the other CCP incapable of injecting into the RCS cannot be performed from the control room. Swapping of CCP trains is a short duration evolution but must be performed in a controlled manner especially when coordinating activities outside the control room. The 4 hour time allowance provides for normal operation of the plant and allows plant manipulations / evolutions to be performed in a time frame in which they can be safely performed.
l l l Am ndmtnt No.103 (Callaway) and Amendmsnt No. 89 (Wolf Crnk) revised current TS 3.5.4 to provide a 4 hour AOT to restore one CCP to an inoperable status in MODES 5 and 6. This 4 hour AOT was specifically reviewed and approved by the NRC as noted in their safety evaluations for those license amendments. This portion of the COMS/LTOP Applicability is the most limiting, as it may involve water solid operation. Current TS 3.5.3 (SR 4.5.3.2) allows 4 hours to secure one CCP after entering MODE 4 from MODE 3. Current TS 3.5.2 requires both CCPs to be operable in MODE 3. Therefore, all of the ITS 3.4.12 Applicability is based on the current TS except for MODE 4 beyond 4 hours after entry from MODE 3. NSHC LS-24 justifies 4 hours for all of MODE 4. ATTACHED PAGES-None I l l 1
l l l ADDITIONAL INFORMATION COVER SHEET l ! ADDITIONAL !NFORMATION NO: O 3.4.13-1 APPLICABILITY: DC,WC REQUEST: Change 6-25 LS-26 (Diablo Canyon and Wolf Creek) i Comment: The change discussion is not adequate. The NSHC contains the necessary justification. FLOG RESPONSE: DOO 6-15 LS-26 has been revised to incorporate additional justification from NSHC LS-26 from Enclosure 4. DOC 6-25 LS-26 has been revised to include: "The RCS is isolated from other systems by valves. During plant life these interfaces can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. Increasing allowed leakage limits from 1 gpm up to 5 gpm for the pressure isolation valves will not challenge the pressure relief capacity of interfacing systems. This amount of leakage is considered negligible when compared with the capacity of the pressure relief valves. Pressure isolation valve leakage limits apply to leakage rates for individual valves. The basis for this LCO is the 1975 Reactor Safety Study (NUREG-75/014)) which identified potentialintersystem Loss of Coolant Accidents (LOCAs) as a significant contributor to the risk of core melt. A subsequent study (NUREG-0677) evaluated various pressure isolation valve configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leak testing of the pressure isolation valves can substantially reduce intersystem LOCA probability. The previous criteria of 1 gpm for all valve sizes is ccasidered arbitrary and is not an indicator of imminent accelerated deterioration or potential valve failure. A study (EG&G Report, EGG-NTAP-6175) concluded allowable leak rates based on valve size was superior to a single allowable value. The single value imposes an unjustified penalty on the larger valves without providing information on pctential valve degradation. In addition, enforcing the single value criteria resulted in higher personnel radiation exposures because larger valves must be repsired in place." ATTACHED PAGES: I I Encl.3A 14
l 1 : 4 l i CHANGE
~
NLDBER HSt[C DESCRIPTION 3 l
- h. 6 25 LS 26 The Operational Leakage LCO has been modified to change the allowed leakage limit for reactor coolant system
- pressure isolation valves for consistency with NUREG 1431 Rev. 1. The RCS pressure isolation valve LC0 permits i system operation in the presence of leakage thro valves
- . in amount
- ; which do not. compromise safety. twsstw 34-l%Jos.4.is-l 6 26 LS 30 The CTS surveillance requirement for performing an RCS water inventory balance is modified to allow deferral of the water inventory balance such that it would be performed in within 12 hours after achieving steady state conditions. The RCS water inventory balance must be i performed with the reactor at steady state conditions as J discussed in the ITS Bases. This change is in conformance j with traveler TSTF 116.
i
,6 27 A RCS leakage detection system descriptions are revised for
! consistency with current TS LC0 3.3.3.1 and USAR Sections l 5.2.5.2.2 and 11.5.2.3.2.2. 4 _- -- 6 28 LG The cur t TS nitio fC LLED Ei delet to consis with REG 1 Rev. . The R seal l er re n flow mit is ved t licen contro ed . ldoc . Se njecti limit ons are stablis by ' ! ECCS bala test pr edures ed from S l j- ) 4.5.2, o a 11 see cont led do nt (ref ence - 1ptG of En sure 3A i the ECCS onversio acka nd i j ttle valv positio urveilla _ in IT_ R _ j (bythet 3.5.2 . JNSEm.T 3A-I4 A.y - _ R4 5.5.5-y ! 7 01 R- Not Applicable to WCGS. See Conversion Comparison Table l (Enclosure 38). ! 8 01 LS 16 This change in conformance with NUREG 1431 Rev.1. revises the applicability of the specification to MODES 1, 2. or 3 with (T,.,) 2 500*F. The change deletes the requirement to
- perform an isotopic analysis for Iodine every 4 hours in
- Modes 4 and 5 and in Mode 3 below 500*F. whenever the i reactor coolant exceeds its Dose Equivalent I 131 limit.
l In addition, this change deletes the requirement to i perform the once per 4 hour surveillance for Dose j Equivalent I 131 in the event the gross specific activity 1- limit is exceeded, in accordance with industry traveler i TSTF 28. The latter is an unnecessary requirement since i l WCGS-Description of Changes to CTS 3M.4 14 $/15M7 1 : _ n __ __ Ih_- 29 t_S. 38 iMSERT 3A-14 f _ G 3.4.14 =1.
]
A NsE C- 3A - 14 d. (6-30
INSERT 3A-14a 0 3.5.5-2 l l The current TS definition of CONTROLLED LEAKAGE is deleted as discussed in DOC 1-28-LG in Section 1.0. The RCP seal water return flow limit is moved to a licensee controlled document. Seal injection limitations are established by the throttle valve position surveillance in CTS SR 4.5.2.g.2) which is moved to ITS SR 3.5.2.7. This surveillance ensures that the ECCS analyses remain valid. Since facility performance and operational details of the type embodied by the RCP seal water return flow limit are required to be described in the USAR per 10CFR50.34. it is acceptable to move the requirements of CTS LCO 3.4.6.2.e and CTS SR 4.4.6.2.1.c to the U3AR. I INSERT 3A-14b 0 3.4.13-1 The RCS is isolated from other systems by valves. During plant life these l interfaces can produce varying amounts of reactor coolant leakage through ! either normal operational wear or mechanical deterioration. Increasing j allowed leakage limits from 1 gpm up to 5 gpm for the pressure isolation I valves will not challenge the pressure relief capacity of interfacing systems. This amount of leakage is considered negligible when compared with the capacity of the pressure relief valves. Pressure isolation valve leakage limits apply to leakage rates for individual valves. The basis for this LC0 is the 1975 Reactor Safety Study (NUREG-75/014)) which identified potential intersystem Loss of- Coolant Accidents (LOCAs) as a significant contributor to the risk of core melt. A subsequent study (NUREG-0677) evaluated various pressure isolation valve configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leak testing of the pressure isolation valves can substantially reduce intersystem LOCA probability. The previous criteria of 1 gpn, for all valve sizes is considered arbitrary and is not an indicator of imminent accelerated deterioration or potential valve failure. A study (EG&G Report, EGG-NTAP-6175) concluded allowable leak rates based on valve size was superior to a single allowable value. The single i value imposes an unjustified penalty on the larger valves without providing I information on potential valve degradation. In addition, enforcing the single value criteria resulted in higher personnel radiation exposures because larger valves must be repaired in place."
,-w'p
i l ADDITIONAL INFORMATION COVER SHEET I I ADDITIONAL INFORMATION NO: Q 3.4.13-2 APPLICABILITY: DC, WC, CA, CP 1 REQUEST: Change 6-26 LS 30 and Difference 3.4-36 (Diablo Canyon, Callaway and Wolf Creek) Comment: TSTF-116 has not yet been approved by the NRC. i FLOG RESPONSE: TSTF-116, Rev. 2 is currently under NRC review. This change provides assurance that the RCS water inventory balance will provide meaningful results. The proposed wording in TSTF-116, Rev. 2 was modified from TSTF-116, Rev.1, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by this traveler. This comment is also applicable to CPSES based on the applicability of JFD 3.4-36. ATTACHED PAGES: ! l Encl. 5A Traveler Status page I Encl. SB B 3.4-81 l 1 l l l l
_ _ _ . _ _ _ _ . . - - . . . _ . . . _ _ - . _ _ . . . _ _ . . . . _ _ _ . _ _ _ _ _ . _ _ _ _ . _ - . _ . _ . . _ . _ _ _ . ~ . _ i INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 i TRAVELER # STATUS . DIFFERENCE # COMMENTS TSTF-26 Incorporated 3.4-32 Approved by NRC. TSTF-27 Revh incorporated 3.4-33. M b5 iib O 34"L'II TSTF-28 Incorporated 3.4-22 Approved by NRC. TSTF 54, Rev. I lacorporated NA CAFP bJ MM*FIE. I.	l TSTF-60 lacorporated 3.4-15 Approved by NRC. TSTF-61 Not incorporated Minor change that is adequately addressed in the Bases. TSTF-87, Rev.h Incorporated 3.4-31 (dhyv4. 6[AJQ /738.3.4mNH4. j j TSTF-93 h lacorporated 3.4-17 {Apfefyv4 6 h k Q /4.7.4. i.{l l TSTF-94 h Not Incorporated NA Retained current TS. [TM 3.v.Ms-( [TS M N 2 findratpV_ f JLag',.,- !Q 3.4.h L TSTF-108, Rev. I Not Incorporated NA LCO 3.4.19 does not apply. TSTF-ll3, RevhI Incorporated 3.4-39 4 3.4.tt-3] TSTF-ll4 Incorporated NA Approved by NRC. TSTF-ll6, Rev[' Incorporated 3.4-36 l 43.4. 83-2.] TSTF-136 Incorporated NA (kpprwvel k3 hAC] /718.3.6det f TSTF-137 lacorporated NA [4pp M 63 AM Q [7A3.#- sof l TSTF 138 Not incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6.. TSTF-151 h incorporated NA /7*4 J.4HW9] TSTF 153 Incorporated 3.4-01 hrwmL by MEQ/72J.f-so9 f TSTF-162 incorporated NA (ppawdbyNK.j73f.J.4-So(.}
'GEGG W D Ineorporated See also Cns 3.4-18 and 3.4 20.19 8 4 3L*2-l h3.4-45f5'+'~F)2.
(WfVr90 '7WF"@ lncorporated 3.4-35 143.4.112\ ilkeGJP"RawO Incorporated 3.4-10 DCPPont h b b hKl3C.3NJ.8 oof I ([WOG-87[ Rah Incorg arated 3.4-47 [ 4 3. 4.83- 4 ) M d' Incorporated 3.4-40 Applicable to Callaway and (Tirrf:- 28Q Wolf Creek only. [d?3 4.l-24
- Incorporated 3.4-49 l4 J.+.L'2.- I l S/158 7
I RCS Operational LEAKAGE B 3.4.13 BASES ACTIONS M (continued) down. This action is necessary to prevent further deterioration 1 of the RCPB. B.1 and B.2 If any pressure boundary LEAKAGE exists, or if unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE l cannot be reduced to within limits within 4 hours, the reactor i must be brought to lower pressure conditions to reduce the I severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 ; within 6 hours and MODE 5 within 36 hours. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary. ; The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely. SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. Primary to secondary LEAKAGE is also measured by performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems. The RCS water inventory balance must be met with the reactor at
~ steady state operating conditions %5F11 ear- CDejAinadessura lM0~1l Therefore, ajNote istaddedl allowing;that this SR is not required A be performed ir, = 3 d 4 until 12 hours of after (or>Uts.4snpanha< , puu ten.\, yunv5u% makeup t; Giants, maka.9 end Letda.an omA ACPse>l iEyction sna, v.cmm 4:1.m
- -J =
(continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 8 3.4 81 S/1/SB7
I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.13-3 APPLICABILITY: CP, WC, CA REQUEST: ITS 3.4.13 Bases LCO c. (Wolf Creek, Callaway, and Comanche Peak) Comment: How is the addition of what does not constitute identified leakage consistent with the definition in ITS Section 1.17 l FLOG RESPONSE: The three categories in the definition of identified leakage do not include leakage outside containment. The added text in the ITS 3.4.13 LCO Bases on what does not constitute identified leakage is unnecessary and ' will be removed.
- ATTACHED PAGES:
i Encl. SB B 3.4-79
_ _ . _ _ . _ . _ _ _ _ _ . _ _ . . _ _ _ _ . _ . . . . . _ . . . _ _ . . . ~ . _ _ _ _ _ _ . _ _ _ . _ . _ . _ _ RCS Operational LEAKAGE l B 3.4.13 BASES LCO RCS operational LEAKAGE shall be limited to:
- a. Pressure Boundary LEAKAGE l
No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further ( deterioration, resulting in higher LEAKAGE. Violation of i l this LCO could result in continued degradation of the i ! RCPB. LEAKAGE past seals and gaskets is not pressure ! l boundary LEAKAGE. ) i l
- b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is l allowed as a reasonable minimum detectable amount that the l
containment air monitoring and containment sunp level l monitoring equipment can detect within a reasonable time ) period. Violation of this LCO could result in continued l degradation of the RCPB, if the LEAKAGE is from the pressure boundary. l l- c. Identified LEAKAGE i Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere l with detection of unidentified LEAKAGE and is well within l 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__. l considered LEAKAGE). fldent . edi _ 31 _ leak fr _ porti s of' 7Q.. 1;a ~ ol {S eas- side ' con neent' _1ch c i sol _.
.he: . 'L ge of is nat~ d' r ~r N '3E ct o',
* ~
( ible ~ he_'Prf~ y Coo' nt rc de ( ontai _ P an.fViolation of thislC0 could result in l cont 1nued degraoation of a component or system. I
- d. Primary to Secondary LEAKAGE throuah All Steam Generators (SGs) I l
i i k l $ (continued) WCGS-Mark-m ofNUREG-1431 - Bases 3.4 B 3.4 79 S/1/S/97
l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.4.14-1 APPLICABILITY: CP, WC, CA REQUEST: Difference 3.4-13 (Callaway, Wolf Creek and Comanche Peak) Comment: What is the justification for restricting the testing to check valves with the addition of the term " check"in three places in SR 3.14-1 and its Bases? All PlVs at a 1 , plant may be check valves however, the addition is not consistent with the "or isolation l valve" part of the first sentence of the SR Bases or with the words of required Action A of ITS 3.4.14. For Callaway and Wolf Creek simple deletion of " check" causes a problem with CTS 4. 4.6.2.2.d and 4.4.5.2.2.d for Comanche Peak. FLOG RESPONSE: CTS 4.4.5.2.2 for Comanche Peak requires surveillances be performed on each RCS PlV listed in Table 3.4-1. The valves listed in this table are not all check valves. All the valves listed are subject to the testing frequency of items SR 4.4.5.2.2.a, b, c, and e. In addition, testing of the check valves within 24 hours of actuation was specifically addressed in item d. This CTS surveillance does not contain a 24 hour test requirement for non-check valve PlVs. The STS equivalent of 4.4.5.2.2 for Comanche Peak is SR 3.4.14.1. However the STS SR does not ' appear to limit the 24 hour test requirement to check valves only. Therefore that portion of the STS surveillance was modified to be consistent with the CTS. The Bases were similarly modified. CTS 4.4.6.2.2.d for Callaway and Wolf Creek is similar to CTS 4.4.5.2.2.d for Comanche Peak. However, CTS 4.4.6.2.2.d for Callaway and Wolf Creek does not specify check valves only (as does the Comanche Peak counterpart). Nevertheless, all the valves subject to CTS 4.4.6.2.2.d in Table 3.4-1 for Callaway and Wolf Creek are check valves, given that the CTS SR wording excludes the RHR suction isolation valves. It was decided that the STS wording should be revised consistent with the wording for Comanche Peak to reference only check valves rather than bring forward the CTS list of PlVs and the RHR suction isolation valve exclusion. ATTACHED PAGES: None l l l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.14-2 APPLICABILITY: DC, CP, WC l REQUEST: Change 6-11 LS-11 (Wolf Creek, Diablo Canyon and Comanche Peak) i Comment: The change justifies isolation by a single valve within 4 hours and the use of check valves as isolations. However, the change does not justify the practice of using a second isolation valve. FLOG RESPONSE: DOC 6-11-LS-11 for Comanche Peak and Diablo Canyon is modified to include the bracketed information form NSHC LS-11. DOC 6-11-LS-11 provides justification for isolation by a single valve within 4 hours, the use of check valves as isolation valves, the use of using a second series isolation valve within 72 hours to isolate a leaking PlV. Comanche Peak and Diablo Canyon take credit for a second series isolation valve to isolate a leaking PlV. The Enclosure 3A description for Comanche Peak and Diablo Canyon of 6-11-LS-11 (in part) provides:
"This change in conformance with NUREG-1431 Rev.1, allows for the flow path to be isolated by one valve within 4 hours and (by a second in series valve] within 72 hours. This change is less restrictive and is ,
acceptable because the first valve has been surveilled as meeting the l same leakage criteria as the inoperable PlV and the small probability of a I failure dMng the 72 hour period.. " For Comanche Peak and Diablo Canyon Enclosure 4 NSHC LS-11 provides additional justification for use two series valves as follows:
"The valve used to isolate the inoperable PlV will be leak tested in accordance with the surveillance requirements. With the successful ;
completion of this leak test requirement, there is sufficient assurance that i a single valve can provide adequate isolation for the following 72 hours. [The requirement to employ a second series isolation valve within 72 hours restores the two valve isolation required by the current TS.] The interval during which only single valve isolation of high-to-low pressure interface is provided, is sufficiently short so as to not involve a significant increase in the probability or consequences of an accident previously evaluated." Wolf Creek has evaluated this issue further and determined that the design of the plant is such that the Reactor Coolant Pressure Boundary only contains two qualified pressure isolation valves in series. Therefore, the bracketed STS Required Action allowing the use of a second series isolation valve to isolate a leaking PlV is not required. ATTACHED PAGES: Encl. 2 4-19 Encl. 5A 3.4-37, 3.4-38 Encl. 58 B 3.4-86, B 3.4-87, B 3.4-88
REACTOR COOLANT SYSTEM OPERATIONAL : EAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to.
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-to-secondary leakage through all steam '
generators net r' ted frc'- *he 9:20ter C00!:nt System and 6-05-A ' 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. B ;;m per 9C pum; COF90LLED LE.^K^.CE :: 9:::ter Cc0!:nt Sy:!:r 6-28-LG ' ~
^'^ r'^ ^f 2225 ; 22 ^:t,2^.d
- f.
- c^r !::E: -10.5 gpm leakage per nominalinch of valve size up to al 6 25-LS (maximum of 5 gpmfat a Reactor Cootant System pressure of 2235 r,20 psig from any Reactor Coolant System Pressure isolation Valve. p0:{ d 6-07 LG
- - Teb': 2.' 1 @ 6-10-LG -
APPLICABILITY: MODES 1,2,3. and 46 6-08-LS-9 ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolationg Valves, reduce t J
rate to within limits within 4 hours or be in at least HOT STANDBY wi.hin the # next 6 hours and in col D SHUTDOWN within the following 30 hours.
/ cuc.3.4-co9 J ~(remote rnanuh jo 3.+,14 -2. }
- c. With any Reactor Coolant S9 stem Pressure isolation Valve leakage greater than the above limit. ::tr the ': 2E :: ret 0 ?: i"'"' 6-11-13 Ht , t "" ' " cure.jisolate the high prossure portion of the affected 7 612-M '
(system from the low pressure portion,within four hours by,use of at least I one 68efedim(mlradeactivated fauf6m1Nih or check valve and within _ 72 hou- yusuR a4 cop 5 clouredMan/al, defac)fvated aurorWatie,6]& v e Arr be in at least HOT STANDBY within the next 6 hours and in HGT-(C,QL,D) SHUTDOWN within the following 43@ hours. "'". On PCS p'^=ur0 4-e-n-u-e : = inen e e p:!g. y % U"~^ jQ3.4.14 3] . "(NEW) With the RHR suction isolation valve interlock function inoperable, isolate the 6-22-M uaffected penetration by use of one deactivatedQalve within 4 hours. , hewwtc, tmsu4h [WC14 009 }
"T0ct pr=: urn !:n then 2225 p !g buuy ter 'her 150 ;5; : 0 !!:::1 6-10-LG Obrr; d!::E : the!! be adjutt0d fr-'he n:ture!!=t pr =ure up te 2225 p !; crum'n; th0 !::Enge to be dir0:t!y prop 0'*!one! te presswe f:""^^.t:' '." '.".2 ^r.; ".2:f p;;^ .
r EBB-PV-8702A/B and EJ-HV-8701 AIB are excluded in MODE 4 when in, or during 6-08-LS-9 transition to or from, the RHR mode of operation. ["Each valve used to satisfy this action must have been verified to meet surveillance' 6-12-M i requirement 4.4.6.2.2. WOLF CREEK - UNIT 1 3/4 4-19 3 ar -uppf CTS 3M.4 _ SMS/97 m+ INSEG.T 4- R a.- g g,g_Q
*ww 145F.iLT 4-l%
_ _ . ~ . - _ _ . . . _ . _ . _ . . _ . _ . _ _ _ . - . _ . _ ~ ~ . _ .. _ _ . . _ 1 l. RCS PIV Leakage 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage l LCO 3.4.14 Leakage from each RCS PIV shall be within limit. APPLICACILITY: MODES 1, 2, and 3, l MODE 4. except valves in the residual heat removal (RHP.) flow path l when in, or during the transition to or from, the RIR mode l' of operation. ACTIONS
.....................................N0TES - - - ---- -- - - - -- ---- --
- 1. Separate Condition entry is allowed for each flow path.
- 2. Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable PIV.
l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more flow paths - -
- NOTE- - - -
with leakage from one or Each valve used to satisfy more RCS PIVs not within Requi_ red _ActionA.1Eg&11eq6$]re E D. limit. Q et W g)must have been verified to meet SR 3.4.14.1 and be in the reactor coolant lQ 3.4.4-2.] pressureboundaryfor g g" g Q@ gpf go t 1 (continued) i t I' WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 37 S/15M7
RCS PIV Leakage 3.4.14 ACTIONS (continued) CONDITION REQUIRED ACTION C W LETION TIME A. (continued) A.1 Isolate the high 4 hours pressure portion of the affected system from the low pressure portion by use of one ,f g w on ,. -- - PS deactivatedguedperp or check valve.
~
{wc 34 #'t M h~ lat . he hi pre _ re! ~ 72~
- 3. g. 4-2._1
..on _of- .
a ectedf emif he0 _ res Po . .;by _of a ggg{gypgj s .clo manual. deactiv _ :automa c, or.c kLvalve. A;2 h Restore RCS~PIV~to 72 hours within . limits. B. Required Action and B.1 Be in H0DE 3. 6 hours associated Completion Time for Condition A not M met. B.2 Be in H0DE 5. 36 hours C; RHR suction isolation C.1 Isolate the affected 4 hours p;g374;ggg valve System penetration by u_se of
> ;;toci;;;r; interlock one W g Hla e c function inoperable. deactivated pd. ., B4PS M valve.
hate menuA !.4 34 WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4-38 S/1SM7
-.~ . - . - - . --. . . - . - ~ - .-. -
RCS PIV Leakage B 3.4.14 l 1 BASES I l LCO RCS PIV leakage is identified LEAKAGE into closed systems l connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. The LC0 PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a ingle allowable value. 6 @c.3.4-ocl Reference # rmits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the nonnal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power. APPLICABILITY In MODES 1, 2, 3, and 4, this LC0 applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation. In H0 DES 5 and 6 leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. c The leakage may have affected system operabilityfgr,u613atin ofd 1pak1ng41offath [o s.4. M 2.1 l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 86 S/1/5/97
i RCS PIV Leakage B 3.4.14 BASES 1 inued) t ljg.SA.14 l A.1 and A.2 The flow path must be isolated by two valves. Required ; Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the_ PIVs and must be within the RCPB[op-tne,Mt@l pressurf porCorr'hfAS d (q3.4. i4.-2. ) Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts the operation with leaking isolation valves. Required Action A.2 specifies that the double isolation _ barrier (es.4.ie2.] of two valves be restored by(closn@ same-titbee-1Talys ttGal.1ffed forAspatTon-ofJrestoring ;;u kiing the RCS PIV toiWithin limits. The 72 hour Completion Time after exceeding the limit alJows for the~ restoration oflthe leaking PIV;to10PERABLE7 status ~. Thisitimeframe considers the time required to complete the Action and the low probability of a second valve failing during this time period. _ ock of bt mmih bu.t is lmle W 4 jO * *"-b (sec. natpp ) B.1 and B.2 If leakage cannot be reduced, the system isolated, or the other Required Actions accomplished, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. This Action may reduce the leakage and i also reduces the potential fcr a LOCA outside the containment. The Jllowed Completion Times are reasonable based on operating eWrience to reach the required plant ccaditions from full l power conditions in an orderly manner and without challenging l plant systems. (continued) WCGS-Mark-up ofNUREG-1431 Bases 3.4 B 3.4 87 5/U5/97
RCS PIV Leakage B 3.4.14 i BASES ACTIONS degraded the ability of the interconnected system to perform
.(continued) its safety function.
A.1 and A.7, The flow path must be isolated by two valves. Required Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB [or the high pressure portion of the system]. Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. Four hours provides - time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts the operation with leaking isolation valves. Required Action A.2. specifies.that.the double isolation
- b'arrier of two valves be restored by closi6'g some 6th'eFN" **
valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time after exceeding the limit l considers the time required to complete the Action and the- ' I low probability of a second valve failing during this time period.
+ (q s.w. utw-l ]
Ih0 72 hour C0iiipleti0n Ilmu afiar caeeedirs the limil diloW f0r the ie5i.0 alivu vi liic luekeny Ely lv CiER5DLt status.*
-Thi: ti :fr;;e conaiuers i.~ne time requirea to complete thT l Actica sad the 1:;. pieb.bility vi e secono valve tailing ^
_durinn thic norind /Deyic;;;r 90te: Inv upT.lons are ~
;"Ouid?d f "0 U.?Q" ired 8.0tiOC A.2. Ihe 5esend Gpti:T.-
'72 50"r resteration) i: :ppr:priate if isel: tic Of =
- d valv ;;;;1d ple:2 the unii in an en:r.:1" zed _
- ndition.)
B.1 and B.2 If leakage cannot be reduced, [the system isolated,) or the other Required Actions accomplished, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 (continued) 10G STS 3 3.4-82 Rev 1, 04/07/95
i l j RCS PIV Leakage l B 3.4.14 BASES l ACTIONS C.l (continued) l The inoperability of the RHR Systs autoclosurc: suction.' isolation valve interlock Godefs the-RHR spetrnJso+et1o3.saseJs Nc.n. coq \ incep;bic of i;;l;tir; 16 rc; pen;; to a high pres;urc_cendition a end picvcating could.allowyinadvertent openingGf th valvcJiit RCS pressures in excess of the RHR systems design pressure. If the "JESysta cut;ci;;urc RHR~ suction isolation ~ valve interlock
~
is inoperable, operation may continue as long A the affected RHR suction penetration is closed by at least one closed manual or deactivated % valve within 4 hours. This Action accomplishe the purpose of the autoclosurc function interlock. k meks.m mu.3e l we.3.4..cca1 SURVEILLANCE SR 3.4.14.1 REQUIREMEKTS _ _ GM M-15 Performance of leakage testing on each RCS PIV@AfoMipVv4 14e used to satisfy Required Action A.1 end llcquircd Kction X.Z is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost. g &
--E o s.% cm q l a Testir.g is to be performed every months, a typical refueling cycle if the plant does not go into MODE 5 for at least 7 days.
The 8__mont Frequency is consistent with 10 CFR 50.55a(g) (W M-olo ( (Ref35 as contained in the Inservice Testing Program, is within the frequency allowed by the American So iety of Mechanical Engineers (ASME) Code. Section XI (Ref. . and is based on the need to perform such surveillances und the conditions that apply during an outage and the potent 1 for an unplanned i transient if the Surveillance were p formed with the reactor at power. g { wca.mc31 d (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 88 S/1/SB7
ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q 3.4.14-3 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.4.14 Actions Notes 1 and 2 l Comment: The adoption of the STS notes (especially #1 which is a less restrictivo change) is not discussed / justified. i FLOG RESPONSE: A new DOC (6-29-LS-38) has been added to include ITS 3.4.14 Action Note 1 to the CTS markup. This note allows separate Condition entry for each pressure isolation valve (PlV) flow path made inoperable by an inoperable PlV. Also new DOC 6-30-A is added to include ITS 3.4.14 Action Note 2 which specifies entry into applicable Conditions and Required Actions for systems made inoperable by an inoperable PlV. ATTACHED PAGES: Encl. 2 4-19 Encl. 3A 14 Encl. 3B 9 Encl. 4 2, new LS-38 i i
. _ . . -. ~ - - - . ~ - . . . . . .- - .- - .-- -. - . ~ . . . . ~ .-
i ! I i REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakege shall be limited to.
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-to-secondary leakage through all steam ^
generators net ::':':f ' :r 'h: "r: ':- Cer'rn' Sy:".:n and 6-05-A i 500 gallons per day through any one steam generator, i i d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System.
- e. S ;;m ; 9C ;ur; COM'90LLEG4:EAKAGE et : Pr :':- Cr 'ent Sy:t:r 6-28-LG l pr - u : Of???S 20 pf;,2nd
- f. 1 :: '::'er '0.5 gpm leakage per nominalinch of valve size up to al 6-25-LS
[ maximum of 5 gpm[at a Meactor Coolant System pressure of 2235 + 20 psig from any Re actor Coolant System Pressure Isolation Valve. :;rr"-d, 6-07-If,
" ' d'r 2.' Y 6-10-LG l APPLICABILITY: MODES 1,2,3, and 4@ 6-08-LS-9 )
ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE. be in at least HOT STANDBY l within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the _ -
above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage gg jeg,ge, & from Reactor Coolant System Pressure Isolation Valves, reduce the leakage i rate to within limits within 4 hours or be in at least HOT STANDBY within the ' iO d' g> next 6 hours and in ml n SHUTDOWN within the following 30 hours.
/ i.uc. ,3.4 -co 9 l - G w atsrmanu h - }C 3.+.14-2 }
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit. : dure th: ': c'::: ::t: : . - 6-ll LS
'":': . '" - ' ' cure.{ isolate the high prossure portion of the affected 7 6-12 M (system from the low pressure portion,within four hours by,use of at least i one Elefed'piipdEhdeactivated bt6s4th or check valve and within_
72 hoursID std 4,stcop6 clasredManwhl, donc)fvated ausorWatic 6]O fMy e"j5rbe in at least HOT STANDBY within the next 6 hours and in HGT _, _(C,,QLD) SHUTDOWN within the following 43hhours. '" On PCS pre :ur:4-ag-ts. _.m
. . - _ _ . . . _ . . _ e_n_n_ _ a., _ . g g yg
. "(NEW) With the RHR suction isolation valve interlock function inoperable, isolate the 6-22-M taffected penetration by use of one deactivatedQalve within 4 hours.
% m nuaj [Wc 3.4-009 )
*T::t pr- u :: !::: then 2225 ;f; but ;rretr-'h:5 = pr; ::: :!'r ::1 6-10-LG O' r ::d !::'r;: d:!! bc :djusted !: 'h: n:tur:! t::t pr:ru : up t
???S pf; Oru- 'n; the !::h ; t: 50 $redy pre; -'!:ne! t0 prerur:
1"r:nt:! !: th :n: 5:!? ;r;;:r.
~
r EBB-PV-8702A/B and EJ-HV-8701 A/B are excluded in MODE 4 when in, or during 6 08-LS-9 transition to or from, the RHR mode of operation, l r "Each valve used to satisfy this action must have been verified to meet surveillance 6-12-M j t requirement 4.4.6.2.2. WOLF CREEK - UNIT 1 3/4 4-19 3 ark-wofCTS3N.4
~
S/2S/97
** tssEe.T 4- R A- g3,4,g423) is w
- 145ER,T 4-l%
l INSERT 4-19a 0 3.4.14-3 ' ** Separate Condition entry is allowed for each PIV flow path. 6-29-LS-38 ; i l INSERT 4-19b 0 3.4.14-3
*** Enter applicable Required Actions for systems made 6-30 A inoperable by an inoperable PIV.
4
..-_ = - . ~. .- -- - . . - - _
l l CHANGE l tGE l NUMBER DESCRIPTION 6 25 LS 26 The Operational Leakage LC0 has been modified to change the allowed leakage limit for reactor coolant system pressure isolation valves for consistency with NUREG 1431 Rev. 1. The RCS pressure isolation valve LC0 permits system operation in the presence of leakage throggtt valves _ in amounts which do not compromise safety. %gf M-l+b).{Qs.4.ls.) 6 26 LS 30 The CTS surveillance requirement for performing an RCS water inventory balance is codified to allow deferral of the water inventory balance such that it would be performed in within 12 hours after achieving steady state conditions. The RCS water inventory balance must be performed with the reactor at steady state conditions as , discussed in the ITS Bases. This change is in conformance with traveler TSTF 116. l 6 27 A RCS leakage detection system descriptions are revised for consistency with current TS LCO 3.3.3.1 and USAR Sections 5.2.5.2.2 and 11.5.2.3.2.2. 6 28 LG The cur' t TS d nitio fC LLED Ei delet to consis with EG 1 Rev. . The R seal { er re n flow mit is ved t licen contro ed , ldoc . Se inject limit ons are stablis by ECCS balan test pr edures ed from S l 4.5.2 oa li see cont led doc nt (ref ence C -- 1 of En sure 3A i the ECCS onversio ackage and by the t ottle valv positio urveilla in IT R __ (3.5.2 . JNSERT 3A-14y -- L{4 5.5.5 21 7 01 R Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 8 01 LS 16 This change in conformance with NUREG 1431 Rev.1. revises the appliccbility of the specification to MODES 1. 2. or 3 with (T.,) a 500*F. The change deletes the requirement to perform an isotopic analysis for Iodine every 4 hours in Modes 4 and 5 and in Mode 3 below 500 F. whenever the reactor coolant exceeds its Dose Equivalent I 131 limit. In addition, this change deletes the requirement to perform the once per 4 hour surveillance for Dose Equivalent I 131 in the event the gross specific activity
- limit is exceeded, in accordance with industry traveler TSTF 28. The latter is an unnecessary requirement since WCGS-Description of Changes to CTS 3M.4 14 S/15/97
-e b6 2A LS. 38 L W SERT 3 A-14 C.
(c. 3 0 g mseer24.a.aj. *3^ W
l INSERT 3A-14c 0 3.4.14-3 6-29 LS-38 Consistent with NUREG-1431, separate Condition entry is allowed for each flow path with excessive leakage from RCS PIVs. Although this specification provides a limit on allowable PIV leakage rate its main purpose is to prevent overpressure failure of the low pressure portions of ' l connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are l degraded or degrading. Each flow path is allowed separate i Condition entry based upon the functional independence of l the flow paths. That is, the required actions to isolate the L high pressure portions of the affected flow paths, in order to protect the connecting low pressure systems, are not l affected by leaking P!Vs in other flow paths. l l INSERT 3A-14d 0 3.4-14-3 l 6-30' A ITS 3.4.14 Action Note 2 which specifies entry into applicable Conditions and Required Actions for systems made inoperable by an inoperable PIV is added to the CTS. This is an administrative change because it only makes explicit l a general requirement that is already implicit in the CTS. I e f wm -- ep=r - re,9 - -
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 9 of 13 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALUWAY NUtBER DESCRIPTION 6-24 Revises ACTION to require going to COLD SHUTDOWN rather than No - Not part of No - The 600 psig Yes Yes M HOT SHUTDOWN with an RCS pressure less than 600 psig. current DCPP TS. action is not part of the current TS. The Operational Leakage LCO has been modified to change Yes No - Leakage limit Yes No - Already part 6-25 of current TS per LS-26 allowed limit for RCS pressure isolation valves, of 5 .5 gpm is already part of Amenhent 66. current TS. 6-26 The CTS surveillance requirement for performing an RCS water Yes No - Already part Yes Yes ' LS-30 inventory balance is modified to allow deferral of the water of the CPSES inventory balance such that it would be performed within 12 current TS. hours after achieving steady state conditions. No - Current No - Current Yes Yes 6-27 RCS leakage detection system descriptions are revised for A consistency with current TS LCO 3.3.3.1 and USAR Sections systems are systems are 5.2.5.2.2 and 11.5.2.3.2.2. applicable. applicable. , The current TS definition of controlled leakage is deleted. No - Not in CTS. No - Not in CTS. Yes, Moved to USAR. Yes. Moved to 6-28 FSAR. LG 3e RCP seal water return flow limit is moved to a licensee controlled d - nt. No. Amen & ent 98/97 Yes - To be No - Amen hent 89 No - Amenhent 103 7-01 This change relocates toe reactor coolant system chemistry relocated to relocated to TRM. relocated to USAR relocated to FSAR R specification from the Technical Specifications to a Equipment Control Chapter 16. Chapter 16. t licensee controlled document. ' Guidelines (ECG). Yes i This change revises the applicability of the specification Yes Yes Yes 8 01 LS-16 to MODES 1. 2. or 3 with (Ty) > 500*F. The change deletes the requirement to perform an isotopic analysis for Iodine every 4 hours in MODES 4 and 5 and in MODE 3 below 500 F. l whenever the reactor coolant exceeds its Dose Equivalent Iodine or at any time the reactor coolant exceeds Gross Specific Activity limits. 6-29 t 145Efx 3B-h _jna,4,i4.,3 j 6-so (^ _
. cce o ,.c...-.:.,s. vrc,os 5,j5,97
- . w INSERT 3B-9a O 3.4.14-3 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 9 Yes Yes Yes Yes Sep rate Condition entry is allowed for f538 each flow path with excessive leakage from RCS PIVs.
ITS 3.4.14 Action Note 2 which specifies f-30 entry into applicable Conditions and Required Actions for systems made inoperable by an inoperable PlV is added to the CTS.
N0 SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) CONTENTS (continued) LS 34............................................... 66 LS 35. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS 36............................................... 68 LS 37,_....... . . . . . . . . . . . . . . . Not Applicabl e { s 16 M G 3.4.14-45 l V. Recurring No Significant Hazards Considerations "TR" TR 2................................................ 70 TR 3................................................ 72 WCGS-NSHCs-CTS 3N.4 2 $/j$j97
- __ _ m . .._ _ - . _ _ _ _ _ _ . _ _ _ _ _ __ _ ..._. _ _ . _ . .._ _._
INSERT 4-38 0 3.4.14-3 IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-38 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS Consistent with NUREG-1431, the LC0 is revised to allow separate Condition entry for each flow path with excessive leakage from RCS PIVs. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been
- performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:
"The Commission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
'2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety."
The following evaluation is provided for the three categories of the significant hazards consideration standards: 2
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change adds a relaxation to the LC0 by allowing separate condition entry for each PIV flow path. Although this specification provides a 1.imit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting 4 systems. The leakage limit is an indication that the PIVs between the RCS ;
. and the connecting systems are degraded or degrading. Each flow path is allowed separate Condition entry based upon the functional independence ,
of the flow paths. Thus, the required actions to isolate the high pressure ; portions of the affected flow paths, in order to protect the connecting low pressure systems, are not affected by other leaking PIVs. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. l , i l l r
-e ~ -w- **
. . - - - . _ - . - . . - . . _ - . - ~ . - - - - - - - . -
1 INSERT 4-38 0 3.4.14-3 I IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS ! l NSHC LS-38 (continued)
- 2. Does the change create the possibility of a new or different kind of accident from any accident'previously evaluated? I l
l The only accidents that are potentially associated with this proposed change, are those related to potential for an interf acing systems LOCA causing a failure of the low pressure portion of a system outside of - containment with the resulting escape of radioactive material. The Required Actions for this LC0 provide for the isolation of the flow path with valves ! that meet the same leakage requirements as the PIVs and which must be within the RCPB or, for DCPP and CPSES. within the high pressure portion of the system. The protection provided for the low pressure system continues to be maintained and is independent of the actions required to protect other flow paths that may also be affected . This change does not introduce any new overpressure accidents and the existing analyses remain l valid. Thus, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not affect the acceptance criteria for any analyzed event. Overpressure protection of each affected low pressure system continues to be provided by leak tested isolation valves which are independent of other flow paths. The margin of safety established by the LCO remains unchanged. Thus there is no reduction in the margin of safety from that previously established. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS-38" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c): and accordingly, a no significant hazards consideration finding is justified. l l 1
1 l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.14-4 APPLICABILITY: WC, CA I REQUEST: Change 6-24 M (Callaway and Wolf Creek) Comment: Cold shutdown rather than hot shutdown is more restrictive however, the discussion does not address the extension of the time from 12 to 30 hours. FLOG RESPONSE: DOC 6-24-M is revised to add the following: j
" CTS LCO 3.0.3 specifies the standard shutdown track Completion Times ,
when Required Actions and Completion Times aren't met as MODE 3 l within 6 hours, MODE 4 within 12 hours, and MODE 5 within 36 hours. i This DOC changes the termination point of the shutdown track in CTS 3.4.6.2 ACTION c from the non-standard MODE 4 with RCS pressure less I than 600 psig in 18 hours to the standard MODE 5 within 36 hours. The cumulative effect is a more restrictive change." ATTACHED PAGES: Encl. 3A 13 l I
CHANGE NUMBER EE DESCRIPTION ; 6 19 TR 3 This change in conformance with NUREG-1431 Rev.1, removes the specific requirement for performing the PIV surveillance prior to returning a valve to service j following maintenance, repair or replacement. Explicit post n.31ntenance TS surveillance requirements have been j deleted because these requirements are adequately ' addressed by administrative post maintenance programs. 6 20 A Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 6 21 LS 35 Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 6 22 H This change adds a new ACTION to isolate the affected RHR penetration within 4 hours if the RHR suction isolation valve interlock function is inoperable. The function of the RHR suction valve interlock is to protect the RHR system from an intersystem LOCA by preventing the RCS hot leg suction isolation valves from inadvertantly opening when the P.CS pressure exceeds the interlock setpoint. Upon failure of the interlock, the current TS permits continued operation for 72 hours for restoration of the affected subsystem. The improved TS requires action within 4 hours to isolate the affected RHR subsystem. Thus the new ACTION decreases the probability of an intersystem LOCA upon the failure of the interlock. This is a more restrictive change and the new ACTION is in LC0 3.4.14 Condition C of the improved TS. 6 23 LS 25 Specification 3.4.6.1 (Leakage Detection Systems) is revised such that the provisions of Specification 3.0.4 are not applicable. This will allow entry into the applicable MODES with only one of the Leakage Detection Systems OPERABLE subject to the requirements of the ACTION statements. This change is consistent with NUREG-1431 Rev.1 and traveler TSTF-60 and is acceptable because of the diverse means available to detect RCS leakage. Ogsaa.T 3A_- gad. - f @ 3.4.1 T - I ) 6 24 H Action c of Specification 3.4.6.2 (Operational Leakage) is revised for consistency with NUREG-1431 Rev. I to require going to COLD SHUTDOWN rather than going to_ HOT SHUTDOWN l with an RCS pressure less than 600 psigdT)W i.54 mer . l ct p uJ.denn regg>reperft. m Rr %-. % l q 3.4.44-4 i WCGS-Description of Changes to CTS 3N.4 13 S/15/97
INSERT 3A-13a 0 3.4.15-1 In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the detection systems is OPERABLE. This condition is superior from a plant safety perspective than imposing a plant shutdown transient under LCO 3.0.3 which could give rise to an initiating event when the plant's leakage monitoring capability is degraded." INSERT 3A-13b 0 3.4.14-4 CTS LC0 3.0.3 specifies the standard shutdown track Completion Times when Required Actions and Completion Times aren't met as MODE 3 within 6 hours, MODE 4 within 12 hours, and MODE 5 within 36 hours. This DOC changes the termination point of the shutdown track in CTS 3.4.6.2 ACTION c from the non-standard MODE 4 with RCS pressure less than 600 psig in 18 hours to the standard MODE 5 within 36 hours. The cumulative effect is a more restrictive change. l l l i h l I .. .__
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.14-5 APPLICABILITY: DC, WC l REQUEST: Change 6-25 LS-26 (Diablo Canyon and Wolf Creek) Comment The justification of the change is inadequate. The NSHC contains the
- properjustification. ,
l FLOG RESPONSE: See the response to Comment Number 3.4.13-1. ATTACHED PAGES: See attached pages for response to Comment Number 3.4.13-1. I l l l l I
l l ADDITIONAL INFORMAKON COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.15-1 APPLICABILITY: DC, WC, CA REQUEST: ITS 3.4.15 and Bases ITS 3.4.15 Required Action E.1 (Callaway, Diablo Canyon and Wolf Creek) Comment: Callaway and Wolf Creek: As written ITS 3.4.15 does not implement CTS 3.4.6.1 as marked up (allowing up to two methods to be inoperable). Specifically, in the ITS as written, with two monitoring methods inoperable TS 3.0.3 would have to be entered as there is no Condition for two methods incperable. Diablo Canyon: ITS 3.4.15 and Bases ITS 3.4.15 Required Action E.1. E.1 Bases state that "With two of the three groups of leak detection monitoring not operable, the two groups will enter their respective ACTION and Completion statements." What in the construction of the ITS supports that statement and more importantly what is the justification for this as the CTS requires 2 of 3 groups of equipment to be operable? FLOG RESPONSE: Given the independence of the three monitoring systems, the plant can simultaneously be in Conditions A and B, A and C, or B and C, but not in all three given Condition E invoking ITS LCO 3.0.3 if all three monitoring systems are inoperable. If ITS LCO 3.0.3 were intended for the simultaneous inoperability of two systems, Condition E (Condition F in I NUREG-1431) would be so worded. This is also supported by ITS page 1.3-1 (second paragraph of the Description section) and Example 1.3-3 which clearly state the plant may be in multiple Conditions at the same time. DOC 6-23-LS-25 is revised to add:
"In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the detection systems is !
OPERABLE. This condition is superior from a plant safety perspective I than imposing a plant shutdown transient under LCO 3.0.3 which could ' give rise to an initiating event when the plant's leakage monitoring capability is degraded." For Diablo Canyon, the Bases statement "With two of the three groups of leak detection monitoring not operable, the two groups will enter their respective ACTION and Completion statements"is deleted. ATTACHED PAGES: 1 Encl.3A 13 l Encl.3B 8 I Encl. 4 52,53 i r
CHANGE NUtBER tLME DESCRIPTION 6 19 TR-3 This change in conformance with NUREG 1431 Rev. 1. removes the specific requirement for performing the PIV surveillance prior to returning a valve to service following maintenance, repair or replacement. Explicit post me'ntenance TS surveillance requirements have been deleted oecause these requirements are adequately addressed by administrative post maintenance programs. 6-20 A Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 6 21 LS 35 Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 6 22 H This change adds a new ACTION to isolate the affected RHR penetration within 4 hours if the RHR suction isolation valve interlock function is inoperable. The function of the RHR suction valve interlock is to protect the RHR j system from an intersystem LOCA by preventing the RCS hot i leg suction isolation valves from inadvertantly opening when the RCS pressure exceeds the interlock setpoint. Upon failure of the interlock, the current TS permits l continued operation for 72 hours for restoration of the affected subsystem. The improved TS requires action i within 4 hours to isolate the affected RHR subsystem. Thus the new ACTION decreases the probability of an l intersystem LOCA upon the failure of the interlock. This i is a more restrictive change and the new ACTION is in LCO l 3.4.14 Condition C of the improved TS. l l 6 23 LS 25 Specification 3.4.6.1 (Leakage Detection Systems) is l revised such that the provisions of Specification 3.0.4 i are not applicable. This will allow entry into the l applicable MODES with only one of the Leakage Detection l Systems OPERABLE, subject to the requirements of the l ACTION statements. This change is consistent with NUREG-l 1431 Rev.1 and traveler TSTF 60 and is acceptable because of the diver _se means available to detect RCS leakage. Qussa7. 3A- a 3 m3- i @.3.4.. I sr - t ) 6 24 H Action c of Specification 3.4.6.2 (Operational Leakage) is revised for consistency with NUREG 1431 Rev.1 to require going to COLD SHUTDOWN rather than going to _ HOT SHUTDOWN with an RCS pressure less than 600 psigdT)wr ira mer pctpuJdern regyreperft. gg g_g q 3.4.M-4 WCGS-Description of Changes to CTS 3M.4 13 5/15/97
INSERT 3A-13a 0 3.4.15-1 In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the detection systems is OPERABLE. This condition is superior from a plant safety perspective than imposing a plant shutdown transient under LC0 3.0.3 which could give rise to an initiating event when the plant's leakage monitoring capability is degraded." INSERT 3A-13b 0 3.4.14-4 CTS LCO 3.0.3 specifies the standard shutdown track Completion Times when Required Actions and Completion Times aren't met as MODE 3 within 6 hours, MODE 4 within 12 hours, and MODE 5 within 36 hours. This DOC changes the termination point of the shutdown track in CTS 3.4.6.2 ACTION c from the non-standard MODE 4 with RCS pressure less than 600 psig in 18 hours to the standard MODE 5 within 36 hours. The cumulative effect is a more restrictive - change. > 1 I l l l i I l l~ ll , _. . - . - - . - .
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 8 of 13 TECH SPEC CHANGE APPLICABILITY NUtBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 6-16 This change removes the requirement for monitoring the Yes Yes Yes Yes LS-14 reactor head flange leakoff system. 6-17 The definition of steady state is moved to Bases. Yes - Yes No - WCGS does not No - Callaway does LG have this not have this definition. definition. 6-18 This change relaxes the requirement for PlV testing No - Not part of Yes Yes No - already in LS-15 following operation in MODE 5. The previous requirement was current DCPP TS. current TS per testing following 72 hours in MODE 5 which is revised to 7 Amendnent 105. days in MODE 5. This change removes the specific requirement for performing Yes Yes Yes Yes 6-19 TR-3 the PIV surveillance prior to returning a valve to service following maintenance. repair or replacement. IST requirements are moved to Section 5 of the improved TS. Yes Yes No - WCGS does not No - Callaway does 6-20 A have this not have this requirement. requirement. This change increases the RCP seal injection flow Conpletion Yes Yes No - See CN 6 No - See CN a6 6-21 LG. LG. LS-35 Time from 4 to 72 hours. with a new added verification that at least 100% of the assumed charging flow remains available. 6-22 This change adds a new ACTION to isolate the affected RtR No - not part of Yes Yes Yes M penetration within 4 hours if the R}R suction isolation current DCPP TS. valve interlock function is inoperable. The leakage detection system specification is revised h Yes No - The non- Yes Yes 6 23 LS-25 that the provisions of 3.0.4_are not aoplicabl Jbe applicabilty of
; 'E eniforvw s tems em.o be r seperatA.a whtt i 4 3 + 85-3 l 3.0.4 is already t of the current gvekg .o.3, ,
5/IS/97 WrG%'-romvraan ramnariwn TaMo- CTS V4.4 _..______._..-_______________________________r-____. . _ _ _ _ _ - _ . _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _
1 IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS l NSHC LS 25 10 CFR 50.92 EVALUATION FOR I TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE I REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS I Specification 3.4.6.1 is revised such that the provisions of Specification 3.0.4 are , not applicable. This will allow entry into the applicable MODES with only one of l the Leakage Detection Systems OPERABLE, subject to the requirements of the ACTION statements. This change is consistent with NUREG 1431 Rev.1 and traveler TSTF 60 l and is acceptable because of the diverse means available to detect RCS leakage. l@An'd_ [sEATt.S-2. l This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted l below: I "The Comission may make a final determination. pursuant to the procedures in 50.91. that a proposed amendment to an operating license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no , significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2 Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety. "
The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Overall protection system performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed. The primary function of the Leakage Detection Systems is to detect significant reactor coolant pressure boundary (RCPB) degradation as soon as practical to minimize the potential for propagation to a gross failure. The TS requires multiple diverse systems to ensure that leakage from a variety of locations and leakage rates can be detected in sufficient time to take measures to place the plant in a safe condition. These system are passive and can not initiate or increase the consequences of an accident. No credit is explicitly taken for these systems in the accident analyses. Entry into the applicable MODEj,while subject to the compensatory measures called n,dt ope rsb n (4 % N o menn q s tem, A.p.,EL4.a lQ3AIS-l[ for et Lo ;30 dWy WCGS-NSHCs-CTS 3H.4 52 SMSR7
l i
- INSERT LS-25 0 3.4.15-1 L in addition, given' the leakage detection diversity, the ACTION for CTS ~3.4.6.1
'is revised to allow continued operation _for up to 30 days under compensatory actions as long as one of_the detection. systems is OPERABLE. Allowing continued plant operation for 30 days, with alternate methods invoked to further monitor RCPB integrity and at least one monitoring system available to
. detect abnormal leakage, is superior from a plant safety perspective'than imposing a plant shutdown transient under LCO 3.0.3 which could give rise to an initiating event when the plant's. leakage monitoring capability is degraded, l
I l i .- o l [. l
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 25 (continued)
)
for in the ACTION statements will not have any effect on the status of the RCPB. The proposed change will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety related equipment assumed to function during an accident situation. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new of different kind of accident from any accident previously evaluated? !
There are no hardware changes nor are there any changes in the method by which any safety related plant system performs its safety function. The method of plant operation is unaffected since the change is based on the acceptability of the W ACTION statement 6 Yn providing compensatory RCPB leakage detection IG8 48' 3 l capability. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change does not affect the acceptance criteria for any analyzed event. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on any margin of safety. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS 25" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c): and accordingly, a J no significant hazards consideration finding is justified. l l l i WCGS-NSHCs-CTS 3M.4 S3 S/15/9/ l I
ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.4.15-2 APPLICABILITY: WC, CA REQUEST: CTS 3.4.6.1 b&c and CTS 4.4.6.1 b&c markups (Callaway and Wolf Creek) Comment: Have the systems been renamed, were the names in the CTS incorrect, or are different systems being relied on in the ITS? FLOG RESPONSE: The same systems are being used. The system names in the CTS have been revised to be consistent with the names used in USAR Section 5.2.5.2.2. DOC 6-27-A already provides this explanation. ATTACHED PAGES: 4 None i 1
)
I l l { i l l l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.15-3 APPLICABILITY: WC REQUEST: ITS Bases Page B 3.4-97 (Wolf Creek) Comment: In the smooth Bases discussion of A.1 and A.2 it should be "and makeup" not "andmakeup" FLOG RESPONSE: The smooth copy of the ITS has been marked to read "and makeup". A l final review of the smooth ITS and ITS Bases is planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases. ( ATTACHED PAGES: None l 1 l l l l l l l-l l
. . . - - - . . . - . . . . . . - .. -._ -... .. .. -... ~. ... - - . - - . ..- - _ . . - . - . .. -.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.16-1 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-39 Comment: TSTF-113 has not yet been approved by the NRC staff. FLOG RESPONSE: See the response to Comment Number 3.4.11-3. ATTACHED PAGES: None i. l l ( i 4 l
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.16-2 APPLICABILITY: WC REQUEST: ITS Figure 3.4.16.1 (Wolf Creek) Comment: In order to be consistent with the ITS LCO and CTS Figure 3.4-1 the units should be micro (p) Curies /gm and not milli (m) Curies /gm as indicated. FLOG RESPONSE: ITS Figure 3.4.16-1 units have been revised to "( Cilgm)". ATTACHED PAGES: Encl. 5A 3.4-49 l l r i i i t
l RCS Specific Activity 3.4.16 [3.+.16-2.} 300 50 t \, i \ l200 i
\\ j UNACCEPTABLE T
i 5150
\ OPIRATION i
! -N k l
l100 N'
\\
! ACCEPTABLE I
! OPERATICN
! 50 8 I i
, 0 i , i j 20 30 40 50 60 70 80 90 100 ; PERCENT OF RATED THERMAL POWER
- FIGURE 3.4.161 (page 1 of 1) l Reactor Coolant DOSE EQUIVALENT I 131 Specific Activity Limit Versus Percent of RATED THERMAL POWER WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 49 $/1587
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.16-3 APPLICABILITY: WC REQUEST: ITS Bases 3.4.16 Applicability (Wolf Creek) Comment: Page B 3.4-103 of the smooth Bases should read "the reactor" not "thereactor" i FLOG RESPONSE: The smooth copy of the ITS has been marked to read "the reactor". A final review of the smooth ITS and ITS Bases is planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases. 1 ATTACHED PAGES: ' None l l 1
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.4-004 APPLICABILITY: DC, CP, WC, CA REQUEST: This item covers the following changes:
- 1. Revise ITS 3.4.14 (and corresponding CTS mark-ups) to reflect that the RHR suction isolation valves from the RCS are remote-manual, not automatic. (Not applicable to DCPP, WCGS, and CPSES.)
- 2. Define an OPERABLE RCP in ITS 3.4.4 LCO Bases as defined in ITS !
3.4.5 and 3.4.6 LCO Bases. (Not applicable to DCPP.)
- 3. Revise the ITS 3.4.9 Bases for Required Action A.3 to match the Bases changes made for ITS 3.4.5 Required Action C.2 per approved ,
traveler TSTF-87 with regard to ways to make the Rod Control System incapable of rod withdrawal.
- 4. Revise the Applicability Bases for ITS 3.4.10 to reflect the change to the Note in Enclosure SA under JFD 3.4-18, i.e., the Note allows entry into MODE 3. MODE 4 should be struck-through as it was in Enclosure SA since the pressurizer safety LCO is not applicable in MODE 4. (Not applicable to DCPP, WCGS and CPSES.)
ATTACHED PAGES: Encl. SB B 3.4-19, B 3.4-43 i l I
RCS Loops-MODES 1 and 2 B 3.4.4 BASES (continued)
~
APPLICABLE Q4.GQ The plant is designed to o mrate with all RCS loops in operlition SAFETY ANALYSES to maintain DNBR above the'11mitivalues, during all normal operations (continued) and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant. RCS Loops-MODES 1 and 2 satisfy Criterion 2 of th; "RC Policy Stet;.;;r,t.10'CFR 50.36_(c)(2)(ii); LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat b the SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power. redh lo seen-* I _ LCdJA **l An OPERABLE RCS loop consisttof an OPERABLE RCPfjp-ophratTonStopt149 forWd flot Wttenspet+Jand an OPERABLE SG_in accordance wTth_ Steam Generator Tube Surveillance ProgramgAn gcp is OPEMASLC
; + ;+ o.s emas,.Ls. ot caim you.><r<6. His akts 4u provids tsread flow . _ + _~ -
APPLICABILITY In MODES 1 and 2, the reactor is critic;l ;r.d thus when: critical has the potential to produce maximum THERMAL POWER. Thui,~To erisu~rithat l the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage. The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for H0 DES 3, 4, and 5. Operation in other MODES is covered by: LC0 3.4.5, "RCS Loops-MODE 3": LCO 3.4.6, "RCS Loops-H0DE 4": LCO 3.4.7, "RCS Loops-MODE 5 Loops Filled"; LC0 3.4.8. "RCS Loops-MODE 5, Loops Not Filled"; LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LC0 3.9.6, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6). (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 19 S/1/S/97
. .. -. . - . . . - - ~.. - - - - - - - _ - .. -. Pressurizer B 3.4.9 BASES LC0 insulation. By maintaining the pressure near the operating (continued) conditions, a wide margin to subcooling can be obtained in the loops, ,,,, _ ., .... m . _ . u ., m, i; dcri;;d fre;; the u;c of ;cacn hc;tcr; r;ted et 17.0 k',1 c;ch]. Thc cr unt accdcd to
;;;; int;in pic;;urc i; dcpcricnt en the tect ic;;;;.
APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, applicability has been designated for MODES 1 and 2. The applicability is also provided 'for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. eMhev- c offMf- In H0 DES 1, 2, and 3 there is a need to maintain the - p a sou.ru.or avail lity of pressurizer heaters, capable of being powered g "" fr emergency power supply. In the event of a loss of offsi e power, the initial conditions of these H0 DES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. For H0DE 4, 5, or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in service, and therefore, the LCO is not applicable. ACTIONS A.1. eM-A.2 A.3 and A.4 Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper limit, even with the plant at steady state conditions. Normally the plant will trip in this event since the upper limit of this LC0 is the same as the Pressurizer Water Level-High Trip. If the pressurizer water level is not within the limit, action must be taken to bring the plant to a H0DE in which the LC0 Joes not apply. resterc the pl;nt to epcr; tion within the bourd; cf the ;;fcty encly;;;. To achieve this status, within.6' hours;the unit must be brought to MODE 3, with all rods; fully inserted ~and incaoable of' withdrawal Additionally, the unit must be brought ( within 12 hours.withThis the takes rc;; torthe tripunit hrc;kcr; out of epca. within 5 heur; and to MOD the applicable MODES. crd rc; torc; the. unit te opcration within the bouri; of th: ;;fcts ;n ly;;;. - lcA 3 4-oouJ. (e.3., by etc energiah on ca.pMa, b3 open n3 tha. gres,er 3 de energia/ng -she enator 3enender sets. )
~
,f (continued)
WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-43 S/1/SR7
4 1 1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 3.4-004, TR 3.4-005, APPLICABillTY: TR 3.4-006, TR 3.4-009 CA, CP, DC, WC REQUEST: Revise the Traveler Status Sheet to reflect that TSTF-54 Rev.1, TSTF-87 Rev. 2, TSTF-136, TSTF-137, TSTF-153, TSTF-162, and TSTF-233 are approved by NRC. Add Rev.1 to TSTF-94 and TSTF-151. , ATTACHED PAGES: Encl.5A Traveler Status page Encl. 68 5 1 l l l l
1 l 1 i INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4-L 1 1 TRAVELER # STATUS . DIFFERENCE # COMMENTS 1 { TSTF-26 Incorporated 3.4-32 Approved by NRC. 5 incorporated TSTF-27 Revh 3.4-33 M ; .3 begit Q Q 1 4.'1.-Il i TSTF-28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev.1 Incorporated FA hypreced-N $3Mme..s.&m9l , TSTF-60 Incorporated 3.4 15 Approved by NRC. 3 TSTF-61 Not Incorporated Minor change that is adequately addressed in the Bases. I i I TSTF-87, Rev.h Incorporated 3.4-31 (ipphywA by MQ /Fle.3.%24j TSTF-93 Q Incorporated 3.4-17 {Aprre'ved63 MAQ /93.4. iJ ] TSTF-94 h Not Incorporated NA Retained current TS. [fR 3.4-Ms-l - j [TS # V f fIng r5or K y N' 14 3.4 E t j TSTF-198, Rev. I Not Incorporated NA LCO 3.4.19 does not apply. TSTF-il3, RevhI Incorporated 3.4-39 q 3,4 i.3} } TSTF-Il4 Incorporated NA Approved by NRC. TSTF-il6, Rev[' Incorporated 3.4-36 l 43.4. 83-2.] TSTF-136 Incorporated NA (kPPrwvU. h 3 M AC]. /T2.3.6do? f j TSTF-137 Incorporated NA [4ppreved.63 AaMg/74 3.v- 3o9 l l j TSTF-138 Not incorporated NA Inconsistent with RCS loops
- requirements of ITS 3.4.5 and l 3.4.6 j TSTF-151 h Incorporated NA /7*4 J.#w291 TSTF-153 incorporated 3.4-01 hrevv4., by NAc]/72 J.9-any /
TSTF 162 Incorporated NA (Appraved by NAC.j7E J.4- So6}
'OEOG M:'Bef"D incorporated See also Cas 3.4-18 and 3.4-20.M 3431*2-I h3.4-45%S)2.
(WfMp90 "IWFa@ Incorporated 3.4-35 [ 4 3.4.nl . 2 \ BheG.69:"RawO Incorporated 3.4 10 DCPP onihbcd h NEC)NI# dof I ([WOG-87[Rev3 Incorporated 3.4 47 L4sA.81-4) M" Incorporated 3.4-40 Applicable to Callaway and (11rrF- 28Q Wolf Creek only. [@3 4.1-2.] h Incorporated 3.4-49 \ 4 3.+.I'2. - I 5/15197
t CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 5 of 8 - SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY ! NtNEER DESCRIPTION 3.4-29 The use of Channel Functional Test (CFT) would be Yes ' No No No l retained from the current DCPP TS to the improved TS. ; 3.4-30 An LCO 3.0.4 exception is added to the Actions of LCO No - This Yes Yes Yes ! 3.4.12. change is out of scope for DCPP. I 3.4-31 Condition C and REQUIRED ACTION D of ITS 3.4.5 and Yes Yes Yes Yes ! Condition A of ITS 3.4.9 are modified to reflect
- generic wording to assure that the rods are fully '
inserted and cannot be wi hdrawn. This change is consistent with TSTF-87 M rut.3.4.cou.) ; In accordance with industry traveler TSTF-26, the Yes Yes Yes Yes 3.4-32 ACTION would be changed to specify taking the plant to a MODE for which the LCO is not applicable. The Frequency of SR 3.4.2.1 is changed to "12 hours". Yes Yes Yes Yes 3.4-33 Ihis change is based on industry traveler TSTF-27. ., 3.4-34 Retains CPSES current TS which requires that the No Yes No No {' precision RCS flow measurement be performed prior to exceeding 852 RTP. Adds a note to SR 3.4.11.1 and SR 3.4.11.2 stating Yes Yes Yes Yes 3.4-35 that the SRs are only required to be performed in . Modes 1 and 2. SR 3.4.13.1 and ACTIONS for LCO 3.4.15 are revised Yes Yes Yes Yes 3.4-36 with the addition of a note per TSTF-116. 3.4-37 The primary to secondary leakage limits are revised No -I No No Yes per Callaway OL Amendment No.116 dated October 1. ( 1996. @ y 4 WCGS-Conversion Conperison Tame-ITS3.4 }. S/lS/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-001 APPLICABILITY: CP, WC l REQUEST: ITS LOC 3.4.8, Note 1.a. is revised to "at least 10 F" consistent with CTS l 3.4.1.4.2. Note 1.a. is a bracketed ([]) note in NUREG-1431, therefore, this change is incorporating CTS inside the brackets. 1 ATTACHED PAGES: l Encl.5A 3.4-18 Encl. SB B 3.4-38 l l l l l l
l l RCS Loops-MODE 5, Loops Not Filled l 3.4.8 l 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops-MODE 5, Loops Not Filled l l LCO 3.4.8 Two residual heat removal (RHR) loops shall be OPERABLE and one RfR loop shall be in operation.
............................N0TES - ---- ---- --- ----- ----
- 1. All RHR pumps may be d; crersind removediihn! operation ? '3.441T/;
for s 15 ;;;inut;; h;n ; witching fice cr.; leap t; ; rett.er 1 _ l hour provided: dM'N m
- a. The[cogoutlet3emperatureligmaintained 10*F EEf8Wf" b_6_1_oiv_is.atur.at. i_ on ~ temper..atur_e_:
gg.
- b. No operations are permitted that would cause a reduction of the RCS boron concentration; and
- c. .; draining agration; to further redu;; th; RCS r.tcr v;1u;;;; er; g.T.itted. Rdactor'yesseliwat_er M3MN, l evel ti siabove' the 3yessel : f1 ange.
- 2. One RHR loop may be inoperable for s 2 hours for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
APPLICABILITY: MODE 5 with RCS loops not filled.
~ ~ik& ~!E*h-Wh* *:& ~ ~: NOTE :. :::fst: - -;---\b t-- ::--
While this LC0]is;not met,' entry into MODE 15 Loops;Not Filled < from MODET51 Loops Filled is~ not persitted. ,3
' O M f[. g .y;;;.;p .g.........{......'..............[..:...e....... .
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop inoperable. A.1 Initiate action to Immediately restore RHR loop to OPERABLE status. l WCGS. Mark-up ofNUREG.1431.ITS 3.4 3.A-18 S/158 7
RCS Loops-H00E 5. Loops Not Filled ; B 3.4.8 i BASES LCO The purpose of this LCO is to require that at least two RHR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via
- the RHR System unless forced flow is used. A minimum of one running RHR pump meets the LCO requirement for one loop in operation. An additional RHR loop is required to be OPERABLE to meet single failure considerations.
Note 1 permits all RE pumps to be d: cr.;rgi ;d: removed;f_ rom I operation for s lh15 ::;i=t;; 2.;r. witd.ir.; fr er.; ';;;; t; onether. The ciretanstances for stopping both R E pumps are to be i arHe limited to situations when the outage time is shor1Tana core _ outlettemperatureismaintained@l0*Fbelowsaturation pac.24.aoi l l ' ,,g,a I o!.P'[8!".." [. ...f M h E h $..$ 9..,.'$.h. w' 5 ' " Y' '5
,,,..... . . . _ . , . . TheiNotegequ. ires pp.s.w.um.g reactoCvessel water]evel'be;above:the1vesselT.fiangeitoTensure the operating RE pump willinot'be intentional _ly deenergized during'ai_diloop;oper_at.1.ons.
Note 2 allows one RHR loop to be inoperable for a period of I s 2 hours, provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be l performed on the inoperable loop during the only time when these tests are safe and possible. An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. IQ 3.4.%2] I APPLICABILITY In MODE 5 with loops not filled MgpMJrticediAPW)s this LC0 requires core heat removal and coolant circulation tFy the RHR System. One RE' loop provi. des .sufficienticapability for this purpose. fHowever, one additional:RE/loopiis'requi. red to be OPERABLE to meet single. failure considerations. Operation in other H00ES is covered by: LCO 3.4.4. "RCS Loops-H0 DES 1 and 2": LC0 3.4.5. "RCS Loops-H0DE 3": LC0 3.4.6. "RCS Loops-H0DE 4": LCO 3.4.7. "RCS Loops-H0DE 5. Loops Filled": LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (H00E 6): and l LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant L Circulation-Low Water Level" (HODE 6). t ! (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 38 S/1/5/97
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-002 APPLICABILITY: DC, CP, WC, CA REQUEST: Clarify ITS 3.4.9 Applicability Bases to state the pressurizer heaters are capable ! of being powered from either the offsite power source or the emergency power I supply. ATTACHED PAGES: Encl. 5B B 3.4-43 l l l
Pr:ssurizer B 3.4.9 BASES LC0 insulation. By maintaining the pressure near the operating (continued) conditions, a wide margin to subcooling can be obtained in the loops. We en;;t d;;ig v;l;; cf [125 k'.? i; derived fre th; a;; cf ;;;;n te;ter; r;t;d ;t 17.0 k" ;;;h). Se ;;nt re;ded t;
; int;in pr;;;;r; i; d;gr.d,.nt en tre te;t 1;;;;;.
1 I APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, applicability has been designated for MODES 1 and 2. The applicability is also provided 'for H00E 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant plap startup. ( eRher c. o(faite- In MODES 1, 2, and 3 there is a_need to maintain the 4' l l' p.w sounA or avail lity of pressurizer heaters, capable of being powered g *~ front emergency power supply. In the event of a loss of offs' e power, the initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. For MODE 4, 5. or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in service, and therefore, the LCO is not applicable. I ACTIONS AJ. end-A.2. A.3 and A.4 Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper limit, even with the plant at steady state conditions. Normally the plant will trip in this event since the upper limit of this LC0 is the same as the Pressurizer Water Level-High Trip. If the pressurizer water level is not within the limit, action must be taken to bring the plant to_a MODE in which theLLC0ldoes not. apply, r;;ter; the pl;nt to egr;ti;n within tre b;;nd; ef th; ;;f;ty ;n:ly;;;. To achieve this status, within 61hoursi_the unit must be brought to MODE 3, with'all rods fully insertediand incanable of withdrawal Additionally, the unit must;be brought ( with tre r;;;ter trip tr;;k;r; egn, within 5 heur; ;nd to MODE 4 within 12 hours. This takes the unit out of the applicable l MODES. ;r.d rc;ter;; th; unit t; e g reti;n within the b;;nd; cf th; ;;f;ty onely;;;. __ Ca.3., 36 de eneryah oil C2.PMs, b 3open.n3 tha. FSs , e - )cA 2 4-oou]. l _ ale enersaing me moto
~
.3 nende, uts. ) ( (continued)
WCGS-Mark-up ofNUREG-1431-Bases 3.4 8 3.4 43 $/1/SM7
- . . . . - _ - - _ _ _ . . . ~ . . _ _ - . - . . . . . _ . _ . . . . .- . - _.. -
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-004 APPLICABILITY: WC REQUEST: Editorial change in ITS Bases 3.4.1 Actions from "C.3" to "C.2". The Required Actions were revised and renumbered and the ITS Bases were not revised accordingly. ATTACHED PAGES: Encl. SB B 3.4-5, B 3.4-6 l
RCS Prsssure. Temperature and Flow DIE Limits B 3.4.1 f-BASES ACTIONS CL.17(continued) WithMfl.owiratefnot:withisi;limitsr thejunitj;isjallowedI21Jours:t.o restoreiflo(ratelto:withinilimits(in accordancelvithlRequirefAction i Cani C T 2 1T C Tr2 T and' Cit 2:3 If3CS~flowfrate31st notf restored itol within ' limitgthelalternative I option 31s.tojreduce: THERMAL P(nER;to < 50*]_RTPrinfaccordanceedth . j Required 1 Action;CT.23Tand: reduce,the Power! Range NeuLtroitf3inxMHigh ! triptsetpoints; tom 55t}RTP'in; accord.ance with]R.equired3W:tten1C3;2:2. . I Redtic]ng (powebto i<iS0r RTP[i ncrease's',th~e:DISTeargin QThe maduction Mn. ! trip setpointsfensuresjthat: continuing operationtremains@tynn ' acceptabl e115 power.:l evel Twith iadequate . DIERimargi nE ..W allowed j Completion:Timeiof72: hours for Required ~ Action)C_3;23:1sjconsistent l withithat.al16wedsfor' Required:Ai: tion C:1:1 and:provideslandicceptable i time to1 reach;.the! required power level.from full] power operation j without; allowing;the plant to= remain in an~unacceptableicondition for an extendediper.iod,,'of time. The CompletionTimes of(2Lhoursifor Required l Actions?C3?lf and C3.2.1~are not additive) The allowed ~ Completion Time;of'6; hours to reset'thejtripisetpoints per Required Action;Cl1l.2.7 recognizes.that, once power;is_.reducedF.the safety analysisjassumptions are satisfied: ~andfadditional timelbecomes available tojreduce:the trip setpoints. If RCS flow ~ rate cannot be restored to withinflimit, the. brought-to .alMODE iniwhich the LCO does not apply.1This: ~ requires the plant mu plant to~be placedlin at least H0DE_2 (RTP s.5%). Once; power _is below 5%, the potential for violating accidentianalysis asstaptions is eliminated. _ fThe. Completion Time of 74 hours for C.1.2;3dsiacceptable because of the increase in the DNB' margin, which is obtained.at0ower power levelse and the low probability of having:a DFBjlimiting event within this time period. C. 2.
- w. 3.4..co 4.]
Subsequent return to: power operation is perfonned in stages to'~ assure that RCS flow rate.is within limits prior to-exceeding 50*:RTP, 75%:RTP and within 24 hours _of ~ achieving :t 95% RTP. ~ Action C;31 assures that
~
the condition. leading;to reduced RCS flow rate has been identified, l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 5 S/1/SR7
RCS Pr:ssure. Temperature and Flow DNB Li: lits B 3.4.1 BASES ACTIONS Scoffinued) yc. 3.4-co4} toithefextentinecessary M ndfcorrectedjpr_forJJ;oJunrestrictedfjouer omation. Tn1s ' Requfred'ActionTis:modifiedibyfa[ Note [ttiatJstatesittiatglERMAL PEER 7doesinotEliave toibe;reducedpigri t o:perfomingithisjAkt:ibnbiFor eXBEP.l.enthis.,neansithargduririgiperformanceloffReiquired"Ast3cnJC'li_17 ifitheIfiMrate isirestoreditoiwithinlimit"atl80%:RTPEMdoesinot needitofbel reduced ~below150%~RTPioru75Y1RTPlto complFlwithlRiqurired Actiori CT2E SURVEILLANCE SR 3.4.1.1 REQUIREMENTS Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits, the 12 hour Surveillance Frequency for pressurizer pressure is sufficient to ensure the pressure can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. SR 3.4.1.2 Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits, the 12 hour Surveillance Frequency for RCS average temperature is sufficient to ensure the temperature can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions. 4 (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 6 S/1/587
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-006 APPLICABILITY: WC REQUEST: CTS 3.4.2.2 establishes the pressurizer safety valve setpoint at 248511% psig. This corresponds to a setpoint pressure range of 2460.15 psig to 2509.85 psig. In NUREG-1431, Rev.1, the setpoint pressure range is identified as 2[2460] psig to s(2510] psig. A portion of the range specified in the ISTS and ITS would be outside the CTS range of 2485 i1% psig. Wolf Creek surveillance procedure, STS MT-005, currently specifies this range as 2461 to 2509 psig. Therefore, the ITS 3.4.10 LCO is modified to reflect the conservative lift setting currently specified in the CTS and plant procedures. ATTACHED PAGES: Encl. SA 3.4-22
l Pressurizer Saftty Valves 3.4.10 3.4 REACTOR C00LAKT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LC0 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift NNS! settings a _ psig and s psig. 24: Z569 ltX1A-64] APPLICABILITY: MODES 1, 2, and 3
= 4 with ;11 "C0 ;;1d 1;; t, gr;tur;; r 275"I. IErf%4E18d
............................N0TE---------------
The lift settings are not required to be within the LCO limits during MODES 3 and+for the purpose of setting the pressurizer @&N383 safety valves under ambient (hot) conditions. This exception is . . i allowed for 54 hours following entry into MODE 3 provided a is$d8M11 , preliminary cold setting was made prior to heatup. ' l l ACTIONS ! C0WITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status. l B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND 08 B.2 Be in MODE 4. with erg 12 hours n,, __, ,__
. m ., -.2 .y Two or raore pressurizer t .g r;turc; s 275'T. g-- - .gyi safety valves inoperable.
I i l
?
( WCGS. Mark-up ofNUREG-1431 -ITS 3.4 3.4 22 $/15197
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-007 APPLICABILITY: DC, WC, CA REQUEST: Revise the Frequency of ITS SR 3.4.11.2 to read: "In accordance with the Inservice Testing Program" consistent with the CTS. ATTACHED PAGES: Encl. 5A 3.4-28 Encl. SB B 3.4-58 Encl.6A 10 Encl. 68 8 1 I
l i Pressuriz;r PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i SR 3.4.11.1 - - - - - - -
- NOTES.
l 1. Not required to be met performed with pg^'gy l l block valve closed in accordance with the i 1-h 2.' Required Action 3of $efxbdprf 64-pv9)[sk OMfrequiredjto.i belperfomed:in M(IlES 1 - I #*~4 I ggggggg
)
l
~ ~
andI2? ! Perform a complete cycle of each block valve {g) 92 days g g gyp I ex. c. -x _ .1. . -.va. .e c . in of _o _ _ ._ _ _~lt J -
.: ) 43. All-4 )
ti ~ -Jor v
..-. ...N0TE__,- --.-- ---- .- _-.._-._- Tn accdam.a. ( J .4 53' mm we.,
Only requiredfto bejperformed'in H0DE51~and.2: ing g u. yg4j3lE :
...................... = ...;..................
u.T'a D$d ' & W SR 3.4.11.2 Perform a complete cycle of each PORV. $$ESEIDIEE l tx 3.4-eb q en e , ,, , es ..,..... n___ ,_ _ _ .____,_
. , . . . . . . . . y,,i,,_ , , . . _ , _. . , .
_ _, ___u __,___
.,... . . . o,, r o,
- 6. 2 - _. _._i. u.. _.
pqg:pge. x s
;ir ;;r.tr;l v;lv; ;r.d ch;;k v;lv; cr, th;
;ir ; c;;;;l;ter; ir, "0"',' ce ,trel y';t; . .
en , , ,, ,
.u ..,.**., i,__u.,..
wwi 7 . na.ni, _ __2 is v a us,w . .wn u , _ _ i.
...,.._________u,O vuivve ui w wwyw.i r, o, La.J >' .-_1'_
n.', eg;psue a -m-~- ______; s___
.s L.wJsury s __ y.www _ _ _ _ _ _ _ _ . . _ _ . . _
e ww u a .uis wesww . ywi rwJ y.www s
".;wi;;3.
l WCGS-Mark.up ofNUREG.1431 -ITS 3.4 3.4 28 5/158 7
I Pressurizer PORVs l . B 3.4.11 l BASES (continued) SURVEILLANCE SR 3.4.11.1 (continued)
'O'4 l REQUIREMENTS _ -- ___
dhattis-__-- _ pableA[bei marydllyicyc ,t maxi Coup ion T1 o res re t PORV a o the bl k va e1 rs, ch is 1w in the lowa e limi (25 to {(7 xtend I test ir block lve F equency s wou be c eted 92 ys. F the the r i of a e, t se r ly c ed bl valve nr oratio of t PORV ERABL tatus ( .e. -e letio of the uir Act (fulfi s the SR) g g. gg g g gg os,4.0-2.} ( b fhe Note.1 modifies this SR by stating that it is not required to l be met performed with the block valve closed, in accordance with i the Required Actions of this LCO. Note:2;nodifiesitMsMto aH ow; entry 11nto rand; operation <1 :37 prior torperfocaing;the SREThis?allowsitheitest to bei reediin MODE?31under operating! temperature'and pres ~ iconditions,' prior'tolentering MW) Ell!or'2: ~ g, g gg gg p 3.4.el-4 l c.a. - ,- .., n. r . n - ! umiSosab l SR 3.4.11.2 {p peg,ic.y Isax e 3 agfrom 4Hs && re 4 i;,epe rs b sik-le. . l SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR.fThe U of M#I 8 hs ifbas n_a typfcal ref0eli V cyc M and nd rd epted 4racti. . . Note modifies t T s SR to allow < entry 31nto i and operation in MODEi3 prior'to; performing the SR;EThisfallJws the~ test to be performed in MODE ~ 3:under operating' temperature ! andjpressure conditions, prior 4to entering-MODE 1lor[2'ZI.n . I accordance.with Reference 5 administrative controls requireithis test;be. performed in'. MODE.3 or(4~to? adequately simulateloperating
, AemperatureandpressureeffectstonPORV' operation. ,- _ -
e l ogerD* a+xFriinca. the reqmM has tweevea. shown htTest these% vaMs usustN *=gae pass ha.SuvW84ws. re 'm Asw l F ,, f __ __ _ _ ex.ccmertsw.e.a. from a _lio.Mt;ct .+ . Np.e. f,t. q m _
= ;4.11.;
Opc; ting th; ;;1;r,eid ;ir ;;ntr;l velv;; ;nd ;ted vel.:.; en th; eir ;;;;;;1;ter; ca;ur;; tra "O~! centr:1 ;y;ta. ;;tu;t;; !~ prepuly .; ten cell;d spen. TPc Ir;qu;ncy of [10] senth; i; be;;d on ; typic;l r;f;;1ing cyci; and th; fr;quercy of th; etter Surv;ille.n;;; u;;d t; d,...en;tr;t; "0"," 0"E",,0!LI"l. S" 2.4.11.4
";i; Sur?;illere; i; ret 7;quir;d f;r pl;nt; With pc;;rir.t IE
! p;; c ;;ppli;; t; tra v;1v;;. j (continued) i'
- WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-58 S/1/SM7
CHANGE Nt#EER JUSTIFICATION 3.4 50 This change is consistent with current TS SR 4.4.9.3.3. The 12 hour frequency applies to vent pathway (s) that are not locked, sealed, or otherwise secured in the open l position. The wording added to ITS SR 3.4.12.5 is also ! consistent with the format used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. 3.4 51 Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 6B). 3.4 52 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). l
.[53 14 6E"T 7*^ lL4C 3 4 * ~l 4
WCGS-Differencesfrom NUREG-1431 - ITS 3.4 10 S/1SM7
l l INSERT 6A 10a WC 3.4 007 l l 3.4-53 This change revises the ITS SR 3.4.11.2 Frequency from "18 months" I to "In accordance with the Inservice Testing Program." The CTS for this surveillance establishes the frequency as being per the IST Program [ CTS SR 4.4.4.1), l l l
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 8 of 8 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.4-50 This change is consistent with current TS SR No - adopting No - adopting Yes Yes 4.4.9.3.3. The 12 hour frequency applies to vent ITS format. ITS format. pathways that are not locked sealed or otherwise secured in the open position. The wording added to ITS SR 3.4.12.5 is also consistent with the format used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. 3.4-51 The Note for SR 3.4.1.4 is removed. This is Yes No No No consistent with DCPP CTS 4.2.3.5. DCPP conducts a measured RCS total flow rate verification on the month frequency. g 3.4-52 Consistent with traveler the Note concerning No - See CN Yes No - See CN No - See CN (Q34 51-2-) accumulator isolation is ved frpm the APPLICABILITY 3.4-45. 3.4-45. 3.4-45. to the LCO. [ -2sD Q M A 6 c.12r % W 'sQ ~ uc. 3.g -co y }
-93_
145E T GB Q.4 IWW9T 6B- 8p[ oc. 3.u..cos ] l l i i WCGS-Conversion Comparison Table-ITS3.4 S/ISi97 .
1 INSERT 68-8a WC 3.4-007 TECH SPEC CHANGE APPLICABILITY i NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY - 3.4-53 This change revises the ITS SR 3.4.11.2 No No Yes Yes l ' Frequency from "18-months" to "In accordance with the Inservice Testing Program." The CTS for this surveillance , establishes the frequency as being~per the IST Program [ CTS SR 4.4.4.1]. ! INSERT 68-8b DC 3.4-003 TECH SPEC CHANGE APPLICABILITY ! I DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER [; 3.4-54 Consistent with the current license bases Yes No No No as approved LA 124/122. ITS LCO 3.4.13 is , revised to reflect reduced steam generator t t primary-to-secondary leakage limits of 150 gallcas per day from any one steam generator and an additional surveillance Requirement to determine primary-to- l secondary leakage every 72 hours. . i t {
)
t _. ._ _. __ _--__--_____--______-___._____.-__-___o-
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-008 APPLICABILITY: WC, CA REQUEST: Revise ITS SR 3.4.12.8 Bases to clarify when the SR must be performed. ATTACHED PAGES: Encl. 5B B 3.4-75 l l
LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.6 (continued) REQUIREMENTS develops excessive leakage or does not close (sticks open) after relieving an overpressure situation. The 72 hour Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open. SR 3.4.12.7 Not Used. Eech icquircd PJC suction rclicf volvc ; hell bc da.enstretcd 0"E"Jf,L by verifying it: PJC ;uction v;1vc ad "JC ;uction isolation valvc arc ; pen and by testing it in exordonn with the In xrvi n T a ting Progren. ("cfcr to 5" 3.4.12.4 for th; "JC
;uction velv; Surveillac; =d for ; dcxription of the requira.mts of the Innivia Tating "regre;;;.' This x Surveillacc i; only pufsrud if the "JC ,uction rclicf valx i:
being und to ;;ti:fy this LOO.
@lackof +c.d mihns M guW is. liric4/drude.mnush q a.4.0en-) \
SR 3.4.12.8 Performance of a COT is required within 12 hours after decreasing RCS temperature to s 275 360*F and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the PTLR allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required. The 12 hour allowance ricqucacy considers the unlikelihood of a low temperature overpressure event during this time. ank Q A Note has been a ded indicating that this SR isWequired to be met performed 712 hours after decreasing RCS cold leg temperature to s275 368*F. 'hc CO' cannot bc perforad until in the L"J" C CS When_thc_P0"" lift x tpoint a n bc redux d to the LT0" nttinv te mus e per rme witAfn J2' hour's pfter] lWC,9A-000 } Q the P S. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 75 S/1/5/97
_ _ - . . . _ ~ ___ _ _- __ ________ _ _ ._, . _ . .- - _ - - . . _ _ . _ _ _ _ l I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-009 APPLICABILITY: WC i REQUEST: Revise ITS 3.4.14, Required Action A.1 and C.1 from " deactivated automatic" to
" deactivated remote manual" since the RHR suction valves are remote manual valves and all other pressure isolation valves are check valves.
ATTACHED PAGES: l Encl. 2 4-19 ! Enc.5A 3.4-38 Encl.58 B 3.4-88 l l 1 l
'h REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE I l LIMITING CONDITION FOR OPERATION l l 3.4.6.2 Reactor Coolant System leakage shall be limited to.
- a. No PRESSURE BOUNDARY LEAKAGE.
l
- b. 1 gpm UNIDENTIFIED LEAKAGE,
, c. 1 gpm total reactor-to-secondary leakage through all steam ., generators no!!: ' !:d frem '50 Per-* - CC !:n Sy: tem and 6-05-A 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, ,
- e. S ;pm p r PC pum; COM790LLED LE.^K^.OE et : 9:20t - CO 'ent Sy: tem 6-28-LG p ::: ': Of 2225 ; 20 ;;;, Ond 4 f. *
=n :M : f0.5 gpm leakage per nominalinch of valve size up to al 6-25-LS
[ maximum of 5 gpmfat a Keactor Coolant System pressure of 2235120 psig from any Reactor Coolant System Pressure Isolation Valve. Opermed 6-07 LG ,
- T:5!: 2.' #
610-LG n APPLICABILITY: MODES 1,2,3, and 46 6-08-LS-9 ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
J
- b. With any Reacter Coolant System leakage greater than any one of the above limits, raciuding PRESSURE BOUNDARY LEAKAGE and leakage 4 e, A j
from Reactor Coolant System Pressure Isolation Valves, reduce the leakage j . rate to within limits within 4 hours or be in at least HOT STANDBY within the iO *b. " b> next 6 hours and in COL D SHUTDOWN within the following 30 hours. ## ( cuc.3.4 -oo9 J -Cremow. rnan uh }o 3.+,14-2 ) I
- c. With any Reactor Coolant System Pressore isolation Valve leakage greater than the above limit. redue the ': S ::::te t: " 611-LS
'Ht t"- ' heure.{ isolate the high prossure portion of the affected 7 6-12-M (system from the low pressure portionwithin four hours by"use of at least i one klefed'orftmlradeactivated WWIb or check valve and within 72 hou seWT aMcop6 closrecManuiiil, dWacjfvated autorWat6FA be in at least HOT STANDBY within the next 6 hours and in HOT {C,0@) SHUTDOWN wit the fon!owing 4@ hours. "" On PCS pr Sture
. 4-
_.a. i. _._ _. .. m... . . e. n. n.
-. ;,. g g s ,21-ts. . "(NEW) With the RHR suction isolation valve interlock function inoperable, isolate the '
6-22-M ~ gaffected penetration by use of one deactivatedQalve within 4 hours. , (reenote rnandah [ WC 3.4-009 \
*T::: pr:::ure:!::: then 2225 pe; but greater 'han le r; :: !!::: d. 6-10-LG Ot: ; d !: E ;; ch:!! be etjutted for h: n:tur:! t :t pre :ure up to a
2225 pri;:: rm'n; th ! h ; ! 50 drect!y pr:pe'tene!! pic Cure d"crenti ! !: th: en: 5:!! pt;;;r. r EBB-PV-8702A/B and EJ HV-8701 A/B are excluded in MODE 4 when in, or during 6-08-LS-9 transition to or from, the RHR mode of operation.
"Each valve used to satisfy this action must have been verified to meet surveillance 612-M L requirement 4.4.6.2.2.
WOLF CREEK - UNIT 1 3/4 4-19 3 ark 9m_of CTS 3M.4 _ S/15/97 m+ IMSEltT 4-t% g g , g .3,}
- wt iOsEft.T 4-L%
_ . _ . _ - . . _ _ . _ _ _ . _ ..__ _..__._._ ____ _._._-__._.._.~___ l RCS PIV Leakage 3.4.14 ACTIONS (continued) C0f0ITION REQUIRED ACTION COMPLETION TIME l A. (continued) A.1 Isolate the high 4 hours ! pressure portion of the affected system from the low pressure l portion by use of one ~ -- 1 (gjestSL.adFW f~~'b_"k""PS l deactivated $5x2 pati) i or check valve. {wc 34-@i M h_ lat heihi 721 l Pre . .. .__lo r h s. +. 8 4-2_] a ected' __ l J , res ,., Po ,. by; e J ffa sggq -l
.j c1o manual deactiv .ed;automa e, or(c valve l.-
A;2 @ Restore RCS'PIV:to 72; hours ! within limits; B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time for Condition A not M met. B.2 Be in MODE 5. 36 hours C: RHR suctionxisolation C.1 Isolate the affected 4 hours valve System ~ ~ ~ penetration by use of O--II*mi wtecl; r a interlock one Med Wantr&e3 _. . function inoperable. deactivated E4 E84PSB4 valve. [ rude.mualt. [ux.3.4.ooi} WCGS-Mark-up ofNUREG-1431 - H53.4
- 3.4 38 S/1SM7
. _ - .m.._- . . _ . _ _ _ . . ..__ _-__ _ _ _ _ _ __ _ _ _ _ _._ _ _
l \ I RCS PIV Leakage i B 3.4.14 i ! I BASES ACTIONS U (continued) The inoperability of the RHR Sy;t;;;; ;ute:1;;;f;<suctioniisolation val.ve interlock (Finders tbs-RMR sp44cnasohrt1oM Nc.a.4. con ir-;; peti; cf i;;leting in r;;p;n;; t; ; high pic;;;r; cendition _ g
;nd pr; Venting could allowlinadvertent openingGf th ;;l;dfi~t ;
l RCS pressures in excess of the RHR systems design pressure. If ' i the ."J'"lSy;te;;; ;ute:1;;ure RIR suctionfi~ solation]alVe interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated {% valve within 4 hours. This Action accomplishe} the purpose of_the sute;1;;ure functi;n interlock. ha e.M m )a { ts.3.4.. coq l SURVEILLANCE SR 3.4.14.1 i REQUIREMENTS _ _ _ qu. H '2.1 l Performance of leakage testing on each RCS PIVMojifipVvg}ve used to satisfy Required Action A.1 ;nd ";quir;; K;tien A.Z 1s-required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per i l inch of nominal valve diameter up to 5 gpm maximum applies to ( each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across l both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be 1 st. reAW
-tos.+.crm-s1 M"Ias.g..om_i)
Testing is to be performed every 18 months, a typical refueling cycle,_ if the plant does not go into MODE 5 for at least 7 days. l The (8_monUi) Frequency is consistent with 10 CFR 50.55a(g) (2 3.5-oio 1 l (Ref @ as contained in the Inservice Testing Program, is within the frequency allowed by the American S iety of Mechanical Engineers (ASME) Code Section XI (Ref. , and is based on the need to perform such surveillances und the conditions that apply during an outage and the potent 1 for an unplanned
- l. transient if the Surveillance were formed with the reactor at l power. hoc 3.g.mcq g
( (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 88 S/1/587
.- . - - - . . . - .. . . . _ . ~ . . - . - . . . . - . . . . . - . . . . . . -.
ADDITIONAL INFORMATION COVER SHEET 1 ADDITIONAL INFORMATION NO: WC 3.4-010 APPLICABILITY: WC, CA REQUEST: Move the CTS list of Pressure Isolation Vcives to the Background Bases for ITS 3.4.14. l ATTACHED PAGES: Encl. 38 6 Encl. 5B B 3.4-85, B 3.4-86, B 3.4-88, B 3.4-90 l . i l l l t I l r I l E 4
CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 6 of 13 TECH SPEC CHANGE APPLICABILITY NUtBER DESCRIPTION DIABLO CANYON CONANCHE PEAK WOLF CREEK Call >WAY 5 02 CTS 3.4.5 is deleted. Steam Generator operability Yes No - CPSES does not Yes Yes A requirements in MODES 1-4 are specified in the RCS loop and have this leakage specifications. specification. , 5-03 Clarification to remove potential interpretation problems Yes No - Same as CPSES Yes Yes A related to probe crientation versus entry point. change 1-15 for CTS Section 3/4.0 6-01 This change adds the performance of an RCS water inventory Yes Yes Yes Yes M balance every 24 hours as a new requirement when the [ containment sump level and flow monitoring system] is inoperable. 6-02 This change allows the performance of an RCS water inventory Yes Yes Yes Yes LS-8 balance every 24 hours as an alternative to the requirement to perform 24 hour samples of the containment atmosphere when a required radioactivity monitor is inoperable. 6-03 This change adds the word " required" to clarify that only Yes Yes Yes Yes A those detectors which are being used to satisfy the LCO must be demonstrated to be OPERABLE. 6-04 The word " DIGITAL" has been deleted to be consistent with No " Digital" not Yes No " Digital" not No " Digital" not A the terminology used in NUREG-1431 as it relates to Channel included in CTS. included in CTS. included in CTS. Operational Tests. 6-05 This change deletes the phrase "not isolated from the No - The phrase is Yes Yes Yes A Reactor Coolant System" when referring to leakage through not part of the i the SGs. current DCPP TS. 6-06 This change moves the LCO for CONTROLLED LEAKAGE (seal Yes Yes No - See CN 6 No See CN 6 A injection flow) from " Operational Leakage" to LCO 3.5.5 LG. LG.
"ECCS" 6-07 This change moves the listing of RCS Pressure Isolation Yes - Moved to the Yes - Moved to Ye_s - Moved to Yes__-Moved t i Valves. FSAR. TRM yr-f6 (ESAFTabW .4-LG IT5 3.414- Bases, lwc. 3.4 -os of WrGS-Conversian Comnariwn Table- CTS 3M.4 S/l587
RCS PIV LGakage l B 3.4.14 l BASES l BACKGROLN) study (Ref. 5) evaluated various PIV configurations to determine (continued) the probability of intersystem LOCAs. 1 PIVs are provided to isolate the RCS from the following typically connected systems:
- a. Residual Heat Removal (RM) System:
- b. Safety Injection System; and
- c. Chemical Volume Control System.
(bel *3 _-_ , PIVs are listed _ _ .jJigtefQ6)S;; tion [ ] I A {nsm.T 3.s.4.ssf M 2 4-084 Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure i system and the loss of the integrity of a fission product ] barrier. ! l APPLICABLE Reference 4 identified potential intersystem LOCAs as a l SAFETY ANALYSES significant contributor to the risk of core melt. The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the Rm System outside of containment. The accident is the result of a postulated failure of the PIVs. which are part of the RCPB and the subsequent pressurization of the Rm System downstream of the PIVs from the I RCS. Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low pressure line would result in a LOCA outside containment and subsequent risk of core melt. Reference 5 evaluated various PIV configurations. leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA. RCS PIV leakage satisfies Criterion 2 of th; =C j'ali;y St;t;.-..t. 101CFR150.'36l(c)(2)(11). I (continued) WCGS-Mark-sp ofNUREG-1431-Bases 3.4 8 3.4 85 $/1/$37
INSERT B 3,4-85 BC 3.4-010 VALVE NUMBER FUNCTION BBV8948 A, B C O SI/RHR/ Accumulator Cold Leg injection BBV8949 A, B, C D SI/RHR Hot Leg Injection BBV001, 022, 040, 059 BIT Cold Leg injection BBPV8702 A. B RHR Normal Suction EJV8841 A, B RHR Hot Leg Recirc Ctmt Isolation EJHVB701 A, B RHR Normal Suction > EMV001, 002, 003, 004 SI Hot Leg Injection Ctmt Isolation l EM 8815 BIT Injection Ctmt Isolation EPV010, 020, 030, 040 SI Cold Leg Injection Ctmt Isolation EPV8818 A, B, C, D ' RHR Cold Leg injection Ctmt Isolation ! EPV8956 A, B, C, D Accumulator Injection isolation j i I I I e f f
l RCS PIV Leakage l 1 B 3.4.14 BASES l LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve , degradation and resulted in higher personnel radiation exposures. l A study concluded a leakage rate limit based on valve size was superior to a ngle allowable value. (tx.34-ocl Reference its leakage testing at a lower pressure differential than between the specified maximum RCS pressure and j the normal pressure of the connected system during RCS operation l (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed , rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power. ! 1 APPLICABILITY In H0 DES 1, 2, 3, ana 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4 valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or ; from, the RHR mode of operation. i In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. ACTIONS The Actions are modified by two Notes. Note 1 provides
- clarification that each flow path allows separate entry into a
- Condition. This is allowed based upon the functional
, independence of the flow path. Note 2 requires an evaluation cf l affected systems if a PIV is inoperable. t The leakage may have affected system operabilityffr)s613nitFn of41pak1n97florfiath
- l o 3.4. M -2. T
, (continued) l VCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 86 S/2/SM7
RCS PIV Leakage B 3.4.14 1 BASES I ACTIONS U (continued) The inoperability of _the RHR Sy;tc;;; autecic;ure suction; isolation v_alye interlock (Findefs tja-RNR sptacn,1so+et1oSvatviD Nc.a.S. coq \ in;;p;bic of i;; latins in rc; pen;; to a high pr;;;;r; condition l nd pr;vcating could allow; inadvertent openingChc volv;fa~t "a*g ; RCS pressures in excess of the RHR systems design pressure. If the , - y - --miestre RfR suction' isolation ~Valfve interlock is inoperable, operation may continue as long as the affected RHR l suction penetration is closed by at least one closed manual or deactivated % valve within 4 hours. This Action accomplishe} the p_urpose of the outecic;urc function interlock, h N8.b2 4 lWC3.4.ccM SURVEILLANCE SR 3.4.14.1 REQUIREMENTS __ O M I4-1I ' Performance of leakage testing on each RCS PIVkc450Mip#v4kre used to satisfy Required Action A.1 nd ",cquir;d Eti5n K.Z is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be 1 st. dh relha lo 3,4,a en _i
-t o s.+. c,m-i 1 a Testing is to be performed every 18 months, a typical refueling cycle The dgif the plant monte) doesisnot Frequency go intowith consistent MODE 5 for 10 CFR at least 7 days.
50.55a(g) M 3.4-otD { (Ref M as contained in the Inservice Testing Program, is within the frequency allowed by the American Society of Mechanical Engineers (ASME) Code. Section XI (Ref. . and is based cn the need to perform such surveillances und the conditions that apply during an outage and the potent 1 for an unplanned transient if the Surveillance were p formed with the reactor at power. [wc3g.acq g _ (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 88 S/VS/97
RCS PIV Leakage . B 3.4.14 l BASES I 1 1 i SURVEILLANCE SR 3.4.14.2 ond "" 3.4.14.3 (continued) RREQUIREMENTS l i design pr;ssur; cf 500 psig. The interlock sitp; int thet I l prevents the valves from being opened is set so the actual RCS l pressure must be < 442 425 psig to open the valves. This j setpoint ensures the RHR design pressure will not be exceeded and 1 the RW relief valves will not lift. The 18 ircnth Frequency is based on the need to perform the Surveillance under conditions that apply during a plant outage. The 18 month Frequency is also acceptable based on consideration of the design reliability (and l confirming operating experience) of the equipment. This:SRJiis ! not required toNperformed'whenithe Rm suction isolation valves;ar_e_ openitoisatisfy, LCO:3;4712.. .
,-u___ em. ___ __2 m _; u.. . __ _,,_.2__ mu_ m ...m__,__..__
[ B 5 Tb db .n T J W4 h survu 5 5 5 b%s vy Ivv 4.b d u a u vvy 5 I sy b 5 Ib # \i ti 1 wu kvb a vdus b functi;n to b; die; bled when using th; "J:", "y-t;; action relief j velves for ;;1d ;;;rpres ur; pret;; tion in ;;;;rd;re; with ! S",2.4.12.7. l ! REFERENCES 1. 10 CFR 50.2. I
- 2. 10 CFR 50.55a(c).
l 3. 10 CFR 50, Appendix A, Section V. GDC 55. l
- 4. WASH 1400 (NUREG 75/014), Appendix V, October 1975.
l 5. NUREG 0677, May 1980. l (a { ) & .) SAR l tx. 5.L-bio \
. ASME, Boiler and Pressure Vessel Code, Section XI.
7 . 10 CFR 50.55a(g). 1 i i l WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 90 S/1/SR7 l i
Attrchm:nt 2 to ET 98-0078 Pag)1 of 4 JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS CTS 6.0 - ADMINISTRATIVE CONTROLS ITS 5.0 - ADMINISTRATIVE CONTROLS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES 1
Attachm:nt 2 to ET 98-0078 P ge 2 of 4 INDEX OF ADDITIONAL INFORMATION l l ADDITIONAL INFORMATION APPLICABILITY ENCLOSED l NUMBER ; 5.1-1 CA NA ' 5.2-1 CA, CP, DC, WC YES 5.3-1 CA, DC, WC YES 5.5-1 CA,DC NA 5.5-2 CA, CP, DC, WC YES 5.5-3 CA, CP, DC, WC YES 5.5-4 CA, CP, DC, WC YES j 5.5-5 CA NA ) 5.5-6 CA NA ' 5.5-7 WC YES 5.5-8 CA, CP, DC, WC YES l 5.5-9 CA, DC, WC YES 5.5-10 CA, WC YES 5.5-11 DC NA 5.5-12 WC YES 5.5-13 DC NA 5.5-14 CP NA 5.6-1 CA, CP, DC, WC YES l 5.6-2 DC NA I 5.7-1 CA, CP, DC, WC YES i CA 5.0-002 CA NA CA 5.0-003 CA, DC, WC YES CA 5.0-004 CA NA CA 5.0-005 CA NA DC 5.0-ED DC NA DC 5.0-001 DC NA
- DC 5,0-002 DC NA j' DC 5.0-003 DC NA DC 5.0-004 DC NA
- TR 5.0-003 CA, CP, DC, WC YES TR 5.0-005 CP NA l TR 5.0-006 - CA, CP, DC, WC YES ]
I WC 5.0-ED WC YES l WC 5.0-001 WC YES j
' WC 5.0-002 WC YES l WC 5.0-003 WC YES WC 5.0-004 WC YES WC 5.0-005 WC YES i
- WC 5.0-006 WC YES i
l l
Attachm:nt 2 to ET 98-0078 Pags 3 of 4 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONAL INFORMATION The following methodology is followed for submitting additional information:
- 1. Each licensee is submitting a separate response for each section.
- 2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.
- 3. If a licensee initiated change does not apply, "NA" has been entered in the index column 1
labeled
- ENCLOSED" and no information is provided in the response for that licensee.
- 4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,38, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. Alicensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as " PLANT SPECIFIC DISCUSSION."
- 5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based 1 on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.
- 6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.
- 7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one or more of the other licensees. These changes are not included in the additional information for the licensee to which the change does not apply, as the changes are only for consistency, do not technical!y affect the request for that licensee, and are being provided in the additional information being l provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment request supplement to be provided later.
l i
l Attichm:nt 2 to ET 98-0078 Page 4 of 4 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued)
- 8. The item numbers are formatted as follows:
l [ Source][lTS Section]-[nnn] Source = Q - NRC Question CA- AmerenUE DC-PG&E WC - WCNOC l CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL" is used for the section number. ! nnn = a three digit sequential number l l l l I
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.2-1 APPLICABILITY: DC, CP, WC, CA REQUEST: STS 5.2.2 b and Difference 5.2-2 Comment: TSTF-121 has been withdrawn for modification, combination and resubmission. Use current ITS. FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. This traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the April 9,1997 letter from C. Grimes (NRC) to J. Davis (NEI), with some exceptions. The FLOG submittals have been revised to incorporate TSTF-258. The latest industry status on TSTF-258 is that the NRC has requested changes to Section 5.7, High Radiation Area. See response to Comment Number 5.7-1 for how the FLOG has addressed the NRC comments on TSTF-258. ATTACHED PAGES: Encl. 2 6-2,6-6,6-17,6-18,6-20,6-23,6-24 Encl. 3A 2,3,5,6,7,9,10 Encl.38 2,5,8 Encl. 4 1, New LS-5 Encl.5A Traveler Status page, 5.0-3, 5.0-4, 5.0-5, 5.0-6, 5.0-10, 5.0-11, 5.0-31, 5.0-37, 5.0-38 Encl.6A 1,2,4 Encl.68 1,2,4,5 1 i i i
l 1 l ADMINISTRATIVE CONTROLS l Unit Staff (Continued) I 1
- b. ^* 'x:' x: ? rrf 0;:--' th '? 5 !- t::x n!:nr ;.t:n ' :! h !-'M :::$ T 01-05-A7 i
' : f T' , ;.'::: the U-"
!: !- " ODE * ,2, 2 :: ', et ' :-' n !Mc- d Sxt ^- rekw e. 4 - G. . j i
ch !' S: !- 5: n.. _' x;r; j
- c. An individual from the Health Physics Groups *, qualified in radiation protecton procedures, shall be on site when fuel is in the reactor; l d. .^.LL CORE ^.LT5 ^7' OMS -h:: M cher d 2nd dired; tr;r !: d L"/ . 0143:A7 l ::t:::in .: d Sz::: 0; ._._: r '!:r::d Ext 0;:n".:: Ur:^-f m nal l '- Fue! t di:n;;t: 5 : n: ::t: xxx- x^ :- ;:--t"r det;; ,
I th :; :d: ;
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_____.,_,,.._-__2.2__ _,._____ ____..m
- f. Administrative procedures shall be developed and implemented to limit the working hours of Unit Staff who perform safety related functions; e.g., Senior Operators, Operators, Health Physicists, Auxiliary operators, and key maintenance personnel. ,
,m _
g _ op,,4 IThe _ unt overti work by U Staff mbe rformin afety / { be li in NRC Stadmept Q E.1-1 C . L .82 ). IW sr.a.T 2. - 2a I
- g. The Superintendent Operations or Manager Operations shall hold a senior reactor operator license. ,
l l l l , 1 r I t l "May be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions. I , WOLF CREEK UNIT 1 6-2 Amendment No. 24,44,81 1 Mark-up ofCTS 6.0 $/1587 l l
I' l INSERT 2-2a 0 5.2-1 l The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime. Any l ? deviation from the above guidelines shall be authorized in advance by the E I Plant Manager 4 or the E Plant Manager's 4 designee, in accordance with , L . approved administrative procedures, and with documentation of the basis for j granting the deviation. Routit:e deviation from the working hour guidelines i l shall not be authorized. I l. l l. t l l l l l l h i l I. l
. - . _ . . , m _ . _.m.. _ . _ . . _ _ . . . . _ _ _ _ . . . . . _ . _ . ~ _ . . . _ _ _ _ _ _ _ . _ _ . _ _ . . . _ _
ADMINISTRATIVE CONTROLS 01-04-LG RINGTlON - S.2.3.* Th: !"_EC 2-!! :_2. . ._ . _2. u _ = t '!:n o._enn. t: : 'n:on;!:nt ; :::!n; 9:r- *--::r, o, e m ,c ur. e _.< uo.r_i__.____ . -- - _. .____ , _ _ . . _ _ . . _ _ , . . _ _ _ _ _ _ . . _ . .__.2_ ......-..._____,_i__.. d- :- 2nd ; r'_! ; n:-!:nt: s____ ! '.er': _t!:_ _.,. _:n6d!n; ?!:n': OM-!!:r de !;n.
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6.2.4 SHIFTTECHNICAL ADVISOR w MWWM.n T ift !! provide technical support to theunwrbuctsusapin the areas of thermal hydraulics, reactor engineering and plant j
-ana is with reaa ti to the safe operation of the Unit.
way_eiserstsw.s swtt.cm.rj k b'IM \ 6.3 UNIT STAFF QUAllFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI /ANS 3.1 1978 with the following exceptions:
- a. Licensed Operators and Senior Operators shall meet or exceed the
- ' qualifications of ANSI /ANS 3.1 1981 as endorsed by Regulatory Guide 1.8. Revision 2.
Net ::: ---Y .for !;;n eunctirn 01-04-LG
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t WOLF CREEK - UNIT 1 6-6 Amendment No. 64344r 54 Mark-up of CTS 6.0 S/lSM P v , .,
i ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) .
- e. Radioactive Effluent Controls Pronram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS Of THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The l program shall include the following elements:
- 1) Limitations on the{ functional caEEMud;:C" .; of radioactive liquid and 0245-A' gaseous monitoring instrumentaten including surveillance tests and setpoint determination in accordance with methodology in the ODCM,
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to : 02-14-M
[10 times the concentration values irU10 CFR Part 20, Appendix B, Table @ler Column 2,{to 10 CFR 20.1001-20.2402.)
- 3) Monitoring, sampling, and analysis of radioactive liquid and -
gaseous effluents in accordance with 10 CFR 20.1302 and with the ; methodology and parameters in the ODCM,
.4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioachve materials in liquid effluents released to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,
- 5) Determination of cumulative --d ;n;:^^:I dose contributions " 02-07-A i from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodoloav and parameters in the ODCM. :t '- ' : : ; ?? i;-Armination ,
' of projected dose contributions for radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,
- 6) Limitations on the[ functional capabilitd;;d"t; and use of the liquid 02-05-Al ~
and gaseous effluent trestraent systems to ensure that the appropriate ~ portions of these systems are used to reduce releases of radioactivity when
. the projected doses in a 31-day penod would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix ! to 10 CFR Part 50.
N *# C M tm
- M dioactive material
- 7) Limitations of the dose rate r(sulting fr released in gaseous effluentsio area yond the SITE BOUNDARY 52-I l (rin5EErlb :: e :::: :::::r:c ta "=? ":n 2c. t::: ' 02-is-AT 9, tr _- ", C -4, M hsttba,Aa.sr.orpva.a. A L-4%.f.tle e a-f a. For noble gases: Less than or equal to a dose rate of 500 rnrem/ year to the whole body and less than or equal to a dose rate of 3000 mrom/ year to the skin, and
- b. For lodine-131, lodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrom/ year to any organ.
WOLF CREEK- UNIT 1 6-17 Amendment No. 42 Mark-up of CTS 6.0 S/15/97
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
- 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from lodine-131, lodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas beyond the SITE BO'JNDARY conforming to Appendix I to 10 CFR Part 50,
- 10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation Z-16-A from uranium fue cycle sources confounino to 40 CFR Part 1Qn g g,g,,3 g
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- '50 Or'ir= d'h: ;!:n' "O pre;== 20'! pr=ide (i) ::;r: .
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- for :rinnm=te! m=M:dng.
- g. Diesef FuelOilTestina Proaram A diesel fuel oil testing program to implement required testing of befN new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance enteria, in accordance with the applicable ASTM Standard. The purpose of the program is to establish the following:
- a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
- 1. an API gravity or an absolute specific gravity within limits,
- 2. a flash point within limits for ASTM 2D fuel oil, 0 2. 2.t A LQ The.preenskdws d Tec.hrdcaR. hc. chb 4 01 N 0 3 Grc >ppu caw. b b bcho\opc.d EHb.aud Coahcis bo3vam. {Q 5.7.- ) j WOLF CREEK UNIT 1 6-18 Amendment No. 12,30,07,101 Mark-up of CTS 6.0 S/ ISM 7
-.. . ~ . . . . . - . - . . ~ . . - - . . - . . - . - - - . ~ . ~ . - . - - - - . - . . . - . - .
ADMINISTRATIVE CONTROLS A.NNUAL REPORTS (C:>ntinued) b,"Z~~ , ' # ,Y'" "_-- ._. y _. .' , #._ ,,._ ,-_.) 03-V@
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ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT l I 6.9.1.6 The Annual Radiological Environmental Operstmg Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the resul's of the Radiological Envwonmental Monitoring Program for the reporting penod The material provided shall ! be consistent with the objectives outlined in (1) the ODCM and (2) Sechonr - IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
"The Annual Radiological Environmental Operating Report shall include the l ~03-07-X" results of analyses of all radiological environmental samples and of all environmental ->"%
radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in a format similar to the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing gdata shall be submitted in a supplementary report as soon as possible. g 4346, A_.3
^^ ' U.^.L "ADIOACTIVE EFFLUENT RELEASE REPORT u .N 6.9.1.7 The Annual-Radioactive Effluent Release Report covering the operation of the unit during the previous calander year of operation shall be submitted-befom[ prior to}May 1 of each yearfin accordance with 10 CFR 50.36a] The report shall include a su.1,. -y of the quantities of radioachve hquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objechves outlined in the ODCM and PCP and
. (2) in wi.'viii- i.e with 10 CFR 50.36a and Sechon IV.B.I of Appendix I to 10 CFR Part 50.
MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operatina statistics and shutdown experie - gq L sect . C.^': #all challFneseJdil failu@y to "- _ the i- WWa,efr D riz shall -
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3 submitted on a monthly basis. - * *- ^- 03 08- )
' ' S. "rir- 927 "'- ; C;..
. ..: - , '^!:i'n;"x. D. C. 20:55, "'. : : ;; !:
'h: ??".C ".:;':::! C5x, no latet .aan the 15 th of each month following the calendar month covered by the report.
WOLF CREEK UNIT 1 6-20 Amendment No.-43,65 Mark-up of CTS 6.0 SMS/97
ADMINISTRATIVE CONTROLS v= 5 " o ^ D!.^t!ON OTECT!OM ""00 ^" 03-10-l.A. J "rrdu = ft ;: :n=: rn:t:n pr' eten thd! be pr; rd rnet^-'
'h: r ;i;r:nt: Of 10 CF 2" 20 :nd thd! 5: :- Trd, ~.dn!dn:d 2nd 2dh:rd t: f-- d! g: :f: :! =Mn; ; renn:! rdtf: ; rut.
612 HIGH RADIATION AREA ~03II-A 6.12.1 Pursuant te Par; :;h 20 202(:X5) of 10 CFR Part 20, in lieu of the .
. m z _ _ _, . --
m__ m., , _ _ _ . .o m e . , _ _, _ _m , n e n e ,_ om ; Ir_--ements of 10 CFR 20.1601] each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radiation ist treater than 100 mrom/hr] .
' but equal to or less than 1000 mR/h at 15 - (12 h ) y cm@tMr(liremee4adie- lQ6,2-1I S n rr r - for 2ny = :: ;M:h 'h0 mf Sr p-'.eWeles shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto sha!! be controlled by requinng issuance of a Radiation Work Permit (RWP).
individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be 4 exempt from the RWP issuance requirement dunng the performance of their assigned duties in high radiationareas with exposure rates equal to or less than 1000 mR/hbt 30 cnMMr(N provided they are otherwise following plant l Q L 2.-l J radiation protection procedures for entry into such high radiation areas. Any individual or
.' group of individuals permitted to enter such areas shall be provided with or ,
accompanied by one or more of the following: 1
- a. A radiation monitoring device which continuously indicates the radiatiori dose rate in the area, or
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring 1 device may be made after the dose rate levels in the area have
- been established and personnel have been made kr.c.i f-;=able of them or
- c. ' An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for
' providing positive control over the activities within the area and shall perform penodic radiation survt illance althe frequency specifie@yJHE Z g ; dgnO-i: O - JJifysielt) .{Q E l-1I 03-II-A' b - - -c ~1n the RWP. n- ; -
7,i 6.12.2 in addition to the requirements of Specification 6.12.1, areas accessible 4 f to personnel with radiation levels greater thar6r'soual tol1000 mR/h at 15 := (1 S M.) 03h*il Af [30 cm (12 in.L-- 'h rdt:n rr :: : frm ny n~':n ;'.!i'h: rItf: ;;n :r:r shall be provided with locked doorspr continuously guardedlto prevent 4movAenaed *-
- 03 20-IE-3 D7-17'_ --ntry, and the keys shallbe maintained under the administrative control of the h A
superviser/gypervismg-Qqon duty and/or health physics supervision. , Doors shall remain locKea except during periods of access by personnel under ' an approved RWP which shall specify the dose rate levels in the immediate work [ o E,2-l1, i areas and t'1e maximum allowable stay time for individuals in that area. In ~
; liee af the stay time specification of the RWP, direct or remote (such as closed-circurt TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures te, provide positive exposure control over the activities being performed within the area.
.5WH hau3er/Cmtrei Roc.m Su{v dor -(wc.5.0-o03]
t' I NOLF CREEK UNIT 1 L23 Amenoment No. -!2. :5. 81 11 ark-hp r>J CTS 6.0 S/IS/97 l
i ADMINISTRATIVE CONTROLS 4 HIGH RADIATION AREA (Continued) For individual high radiation areas accessibl_e to personnel with radiation 6' levels of greater than 1000 mR/ Mat 30 cm(W)DfBthat are located within large 03-11-A areas, such as PWR containment, where no enclosure exists for purposes of MRd locking [or that is not continuously guardedjand where no enclosure can be 03-20-LS-3 reasonably constructed around the inomdual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device. d 03-12-LG ,'- su:.a l Ch: ;:: :thePCP-s
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_ _ - . A:??A????? 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)
- - r .a The ODCM shall contain the methodology and permeters used in the W ^
] calculation of offsite doses resulting from radioactive gaseous and liquid effluents,in the calculation of gasnus and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and 1 b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should 4 ..
'-,w[
be included in the Annual Radiological Environmental Operating and Radioactive t - Effluent Release Reports required by Specification 6.9.1.6 and Specification 6.9.1.7. Changes to the ODCM:
- a. Shall be documented and records of reviews performed shall be retained = :: ! :d 5'; C;=ft:S - SMO 2.0. This 83-09-123' documentation shall contain: Jwh 4
j4 1) Sufficient information to support the change together with [ the appropriate analyses or evaluations justifying the change (s) and
- 2) A uetermination that the change will maintain the leve radiocctive effluent control required by 10 CFR 20. 40 02-09-A P '
CFR Part 190,10 CFR 50.36a. and Appendix I to 10 C art - 50 anri not adversely impact the accuracy or reliability of effluent dose, or setpoint calculations.
- b. Shall become effective after :=:;; and ac=pten= by the F;PC and 02-13-LG the approval of the Plant Manager.
- WOLF CREEK UNIT 1 6-24 Amendment No. 42.15. ES,100 3fark-up of CTS 6.0 5/15/97
I CHANGE NUMBER HSBC DESCRIPTION performed in accordance with guidance in an NRC letter dated October 25, 1993, from William T. Russell to the i chairpersons of industry owners groups. ' f Furthermore, sections which contain details of procedure review and approval, including temporary changes, are contained in regulations and standards (10 CFR 50.36; 10CFR 50 Appendix B, Criterion II and V: ANSI N18.7 1976: and N45.2 1971). Therefore, duplication of these requirements in the CTS is not required. 01 05 A The requirement for the presence of a Reactor Operator ! (RO) or an SR0 in the control room is deleted from the TS l since the requirement is consistent with and duplicative l of the manning requirement in 10 CFR 50.54(m)(2)(111). l l Deletion of the CTS requirements does not change the , i manning requirements and is therefore considered an I administrative change. L 01 06 LG The details regarding the minimum shift crew requirements l have been removed from the CTS because they are redundant l to 10 CFR 50.54(k), (1), and (m) with the exception of the requirement for non licensed operators. The corresponding ITS section 5.2.2b requires meeting the requirements of these regulations which specify the shift complement regarding licensed operators for all modes of operation. I The minimum shift crew requirements will be moved to a licensee controlled document.
]
01 07 LG Revises Section 6.2.h , Unit Staff Organization, to reflect the non licensed operator staffing requirements for a single unit site consistent with the NUREG 1431, Rev.1 requirements. The minimum shift crew composition as described in Table 6.21 has been moved to a licensee controlled document and provides requirements for the minimum number of non licensed operators necessary for plant operations. This proposed change is consistent with NUREG 1431 Rev. 1. 01 08 LG Move the fire brigade requirements to a licensee controlled document. Moving these requirements is consistent with NUREG 1431. Rev. 1. These requirements can be found in BTP ASB 9.51 and their duplication in the l ITS is not required. l __ ~-_ _ _ I 01 09 ) lot" ped 8.INSEP_ _ _T 3A - g '~cp s. 2. - l { (A i WCGS-Description of Changes to CTS 6.0 2 S/15/97 l
1 INSERT 3A 2a 0 5.2-1 l 1-09-A The CTS requirements concerning overtime being in accordance with ! the NRC Policy Statement is replaced by referring to administrative procedures for the control of working hours. The , proposed change provides reasonable assurance that safe plant ! operations will not be jeopardized by impaired performance caused by excessive working hours. Specific controls for working hours l of reactor plant staff are described in procedures that require a deliberate decision making process to minimize the potential for impaired personnel performance, and that established procedure , control processes will provide sufficient controls for changes to ! that procedure. Replacement of the CTS reference to referring to administrative controls does not change the requirements associated with working hours and is therefore considered an administrative change. These changes are concistent with the NUREG-1431 as modified by TSTF-258. l l l 4 l l i l 1 ( - -
~ _ _ . _ _ . . . __ _ _ . _ . . . _ . _ . -. _ _ _ _ _ _ _._._ __ _ _ _
og . w 'd r4et apyhcalAA.ie LAC 6 5. See. Convesm Compavi.h i2Q ( Gnetoswe. so , - j
' ~
CHANGE 78. 5.o-ooT I NLNSER NSiC DESCRIPTION l 01 10 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 01-11 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 01-12 A . Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 01-13 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 01-14 A Not applicable to WCGS. See Conversion Comparison Table _r h5 02 01 A A Wserr SA-3 (Enclosure g 38). .l e s 2-lj CTS Section [6.6.la] for Reportable Event actions has been deleted from the CTS. This section only repeats the regulatory reporting requirements defined in 10 CFR 50.72 and 10 CFR 50.73, and is unnecessary in the TS. Deletion of this section from CTS does not impact safety because it is redundant to the regualtions cited and is therefore acceptable. i- 02 02 LS 4 CTS Section [6.7], " Safety Limit Violation," requirements to notify the NRC within 1 hour following a violation of a safety limit (SL). submit a Safety Limit Violation Report 1 and not restme plant operation until authorized by the . Connission are being deleted. These requirements are a 1 duplication of 10 CFR 50.36(c)(1),10 CFR 50.72 and 10 CFR 50.73. [The 14 day Safety limit Violation report in the CTS is not required since 50.73 would require 30 day Licensee Event Report.] Since the plant must meet the applicable requirements contained in the regulations, sufficent regulatory controls are maintained to allow removing these duplicate regulatory requirements from the current TS. The notification requirement to executive management and the review committees is an after the fact notification and is not necessary to assure safe operation of the facility. As such, this requirement is not necessary to be included in the TS. These changes are consistent with NUREG-1431 and traveler TSTF 5. 02 03 A The implementation procedure requirements related to [the security plan, the emergency plan,] process control programs and radiological environmental and offsite dose calculation programs are deleted from the CTS consistent
- with NUREG 1431. These types of procedures are either required by Regulatory Guide 1.33, Rev. 2. Feb.1978 WCGS-Description of Changes to CTS 6.0 3 S/158 7
INSERT 3A-3a 0 5.2-1 1-15-A This change revises the CTS to eliminate the title of " Shift Technical Advisor (STA)." STAS are not used at all plants (the function may be fulfilled by one of the other on-shift individuals). This Section is revised so that it does not ; imply that the STA and the Shift Supervisor must be different individuals. Option 1 of the Commission Policy Statement on Engineering Expertise on Shift is satisfied by assigning an individual with specified educational qualifications to each operating crew as one of the SR0s (preferably the Shift Supervisor) required by 10 CFR 50.54(m)(2)(i) to provide the technical expertise on shift. Eliminating the title of STA is considered an administrative change since the requirement for engineering expertise on shift is maintained. This change is consistent with NUREG-1431 as modified by TSTF-258. l d l
CHANGE NUMBER tlSlE DESCRfPTION 02 09 A The description of the [0DCM] (or equivalent programs and procedures) was revised to be consistent with NUREG 1431. The [0DCM] description is also revised to reflect new 10 CFR Part 20 requirements. 02 10 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 02 11 M New program requirements " Safety Function Determination l Program" and " Bases Control Program" would be added, consistent with NUREG 1431. Although these new programs reflect current plant practice, delineating them in the ITS would be more restrictive. 02 12 LG This change moves the Emergency Diesel Generator Reliability Program requirement to a licensee controlled document. Moving this. program is consistent with NUREG-1431. 02 13 LG Revises Section 6.14 item b to move the requirement that ODCH (or similar programs and procedures) changes require i review and acceptance by onsite review cosmiittees to the l ODCM. The onsite review of ODCM changes is currently required per [ procedures]. This change is consistent with j NUREG 1431.
- 02-14 M Per GL 89 01, concentrations of radioactive material releases in liquid effluents to unrestricted area shall conform to 10 times the concentration values in Amendix B. Table 2, Column 2 of 10 CFR 20.100120.2401g.
(at :mefirrpsnowii. m eer se-sa _ ' qu.t y g _ l 02-15 LG CTS Section [6.6.1.b] contains requirements for the plant review and submittal of a reportable event. This information is to be unved to a licensee controlled L document. Given that these reviews and submittal of l results are required following the event without a specified completion time, the requirements are not necessary. The moving of this information maintains consistency with NUREG 1431. Rev. 1. L l 02 16 A To maintain consistency with the Bases for TS 3/4.8.1, change the Diesel Fuel Oil Testing Program description for sampled properties of new fuel oil from "within limits" to
" analyzed" within 30 days following sampling and addition of the fuel oil to storage tanks. This wording more 7 clearly defines that within 30 days following the initial new fuel oil sample, the fuel oil is analyzed to establish i
( WCGS-Description of Changes to CTS 6.0 5 5/15/97 l i _ _ -
INSERT 3A-Sa 0 5.2-1 This limitation provides reasonable assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section II.A design objectives of Appendix I to 10 CFR Part 50 and (2) restrictions authorized by 10 CFR 20.1301(e). These changes are consistent with NUREG-1431 as modified by TSTF-258. I t
CHANGE-NUMBER j$t!C DESCRIPTION that the other properties specified in Table 1 of ASTM D975 81 are met. This change is consistent with the Bases for SR 3.8.3.3 of NUREG 1431. 02 17 LS 1 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). l 02 18 A Revises the Radioactive Effluent Controls Program dose rate limits released to areas beyond the site boundary to reflect new 10 CFR Part 20 requirements.%n isten wTtT
@Clet rd 7 8/95 ( risto I. Gr mes t Owne Gro . it nally, MLC sued a raft ri ter - 1 which oposed anges t the S andar [
,Tec cal ifica ons. s cha is co ist wi j Generic etter i
dra NUREG 1 1 Re . 1 - - . , - l by a oposed anges onsi veler t reflect enhith / 10 FR Part . lyaEf.T 34 -(,e j 02 19 LS 2 Consistent with NUREG 1431, the surveillance interval for verifying that other properties are with limits for ASTN 20 fuel oil is changed from "within 30 days" to "within 31 days" after obtaining a sample. The fuel properties that can have an immediate detrimental impact on diesel combustion, (i.e., API gravity, kinematic viscosity, flash point and appearance) are verified prior to addition to the storage tank. The "other properties" may be analyzed after addition to the tank. The 31 day verification interval for these properties is acceptable because the fuel properties of interest, even if they are not within their stated limits, would not have an immediate affect on diesel generator operation. The CTS 30 day verification interval was probably chosen because it was a convenient time interval for sending the sample and receiving the results from the laboratory selected for testing. NUREG-1431 has selected a 31 day testing interval. The 1 day increase in the interval would not have a significant affect on the acceptability of the diesel fuel oil. 02-20 A Consistent with NUREG 1431. Rev. 1 and traveler TSTF 118. add the statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program frequencies. These sentences provide consistency with the current application of these requirements as provided in ISTS 5.5.6 and ISTS 5.5.11. Amendment [101] moved the Diesel Fuel Oil Testing Program to Section 6.0 of the current TS but did not include a statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable. SR 3.0.2 and SR 3.0.3 are applicable to the surveillances WCGS-Description of Changes to CTS 6.0 6 5/15/97 l
l ! i, 1 l l INSERT 3A-6a 0 5.2-1 L After issuance of Generic Letter 89-01, 10 CFR 20 was updated. The NRC
. issued a draft Generic Letter. 93-XX, on proposed changes to STS NUREGS basedlon- the new 10 CFR 20. The proposed changes are consistent with the draf t-generic letter, the April 9,1997 letter from C. Grimes to J. Davis -
(with'some exceptions). The proposed changes maintain the same overall level of effluent' control while retaining the operational flexibility that exists with current TS under the previous 10 CFR 20. These changes are L . intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements. The proposed changes are consistent with NUREG-1431 as modified by TSTF-258. For DCPP and CPSES, portions of TSTF-258 were adopted that were not already incorporated into the CTS based on previous license amendments. I I' l
. 2.3 Apr -td w_6c.S u, con w s co h dm hbO_
CHANGE NUMBER _ MSE __ DESCRIPTION W5ditT 3A ~1@l wc 5 o-co4d which reference these programs, and therefore, the lack of an applicability statement in the Programs introduces confusion. A _ iusER":r .3A -70 ! o s.2. - I ] 03 01 A Revises " Routine Reports" section to be consistent with NUREG 1431. The method for submitting all reports is revised to be in accordance with 10 CFR 50.4. Since this change merely makes the TS consistent with the regulations, it is considered administrative. 03-02 A The requirement to submit a Startup Report is deleted from the CTS to be consistent with NUREG-1431. This report required no staff approval and was submitted after the fact and is therefore not required to ensure safe plant operation. The approved 10 CFR 50, Appendix B. QA Plan, and USAR startup testing program provides assurance that the affected activities are adequately performed and that appropriate corrective actions, if required, are taken. 03 03 A The Annual Reports section is revised to be consistent with NUREG 1431 and traveler TSTF-152. Names and formats are revised consistent with NUREG 1431. Also, revises the annual report section to reflect the new 10 CFR Part 20 requirements ':nd associated recommended changes noted in NRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10CFR20 and 50.36a Changes." (From Christopher I. Grimes to Owners Groups Chairs). 03-04 A The requirement to report specific activity limit violations is deleted consistent with NUREG 1431. This report is a history of Reactor Coolant System (RCS) specific activity Limiting Conditions for Operation (LCO) entries. GL 83-43 and revised reporting requirements in the regulations intended that LC0 entry reports no longer be required. The reporting requirements in regulations cover situations such as seriously degraded barriers (fuel failure). Therefore, every violation of the RCS specific activity LC0 need not be reported. Serious degradation of a fission product barrier, among other more seriots events are required to be reported by 10 CFR 50.73. This c.hange is administrative in that it only affects reports and do not affect plant operations. 03 05 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 03 06 A CTS [6.9.1.7] " Annual Radioactive Effluent Release Report" and CTS [6.14.c] is revised consistent with NUREG-WCGS-Description of Changes to CTS 6.0 7 S/15/97
INSERT sA-7a 0 5.2-1 02-22 A The Radioactive Effluents Controls Program is revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3 are applicable to these activities. These statements of applicability clarify the allowance for surveillance frequency extensions and allowance to perform missed surveillances. Generic Letter 89-01, " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications and the Relocation of Details of RETS to the ' Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents i Control Program in the Administrative Controls Section of the Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 89-
- 01. This change is considered an administrative change since the changes are in the presentation method only. This change is consistent with NUREG-1431 as modified by TSTF-258, i
INSERT 3A-7b WC 5.0-004 02-21 A - Amendment No.106 for Wolf Creek incorporated a footnote to allow the volumetric and surface examination of the RCP "D" motor .. , flywheel for the first 10-year ISI interval be delayed for one l operating cycle. The examinations are completed during the ninth l refueling outage. Since the footnote is a one-time exception and i has been ' satisfied, the footrote is no longer applicable and can be deleted. l l i i l I I
j NUMER NSBC DESCRIPTION i l required by 10 CFR 20.1101(c). The CTS is redundant to requirements in the regulations and thus is deleted.
'03 11' A IThe Hi adiation a is revi to ensiste wi 1431 and new Part requ ts C re
- non techni and add e ifica n and co orm ith
- and RG 8 .8. Wsr,s.T3A-g (NUREG- ,
03 12 LG The Process Control Program (PCP) section is proposed to - be moved outside the CTS consistent with NUREG 1431. The PCP implements the requirements of 10 CFR 20,10 CFR 61.
- and 10 CFR 71. Therefore, relocation of the description of the PCP from the CTS does not affect the safe operation of the facility. The PCP will be adequately described in 3
licensee controlled documents. l 03 13 M _ The following report [s] will be added to the ITS
- Administrative Controls section
- " Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)" [and
" Post Accident Monitoring (PAM) Report."] The PTLR is more restrictive because it is not currently required.
This change is also described in the description of a changes for CTS Section 3/4.4 (PTLR). [The PAM Report is
- already required per TS 3/4.3.J.
1
- 03-14 M Shutdown Margin values would be moved to the Core l Operating Limits Report (COLR) traveler TSTF 9. In i
addition, moderator temperature coefficient limits would also be moved to the COLR. The addition of these
- specifications to the COLR is considered to be more restrictive.
i j 03 15 M Not applicable to WCGS. See Conversion Comparison Table . (Enclosure 38). ! 03 16 A Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 3B). [ '03 17 A Deletes the methodology section references in the COLR. These references are adequately defined by the analytical
- methods themselves as approved by the NRC and it is 4 redundant to repeat the information in the ITS. This
, change is consistent with NUREG 1431. (Q 5.7.-I \ 03 18 CTS 6.9. . " Mon y Opr g Repo is re ed
.N 9 inco ate t reportin of all a11enges nd fa res i to PORV or pres izer sa y valve . The i
uir to r t these allenge nd fai res afas IN SE AT SA-9 a WCGS-Description of Changes to CTS 6.0 ' 9 S/15/97 i
f INSERT 3A-9a 0 5.2-1 03-18-LS-5 The CTS requirement to provide documentation of all challenges to the PORV's or safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694, "TMI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states: l
" Assure that any failure of a PORV or safety valve to close will :
be reported to the NRC promptly. All challenges to the PORVs or l safety valves should be documented in the annual report." NRC ; Generic Letter 97-02, " Revised Contents of the Monthly Operating l Report" requests the submittal of less information in the monthly i operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent with NUREG-1431 as modified by TSTF-258. INSERT 3A-9b 0 5.2-1 CTS 6.12, which provides high radiation area access control alternatives pursuant to 10 CFR 20.203(c)(2) has been revised as a result of the change to 10 CFR 20 and the guidance in Regulatory Guide 8.3.8. Since the plant requirements remain the same, except as identified in specific Description of Changes, the change is ; considered administrative. This change is consistent with NUREG-1431 as modified by TSTF-258.
CHANGE NUMBER DESCRIPTION NSEC
~~ pg,g.i }
E A a s.1.-0 03 19 A The 'rm '" uthor ed" i cha to "in vert t" in thh R ation ea s ion prev ion in erten ntry di ssed in ion .5 of 8.3.8. 4 , is - efle t C*s posi 'on re ding sical b 1ers hi adiation eas. diati areas ithin the l' s1 ed shall lock r con uous1 guar
\t reve inadverte entry discus in 8.3 Furth re, the stinctio twee unaut riz ers ;
i vertent importan ased o a No e of ola " n that Cal ay receiv on this inte etati n of erms. H& us.4 _ 03 20 LS 3 Consistent with NUREG 1431, the use of a continuous guard is provided as an additional option for preventing ' inadvertent entry into high radiation areas that are accessible to individuals. This option is acceptable because it provides access control to these areas comensurate with the other methods already permitted by the CTS. (i.e. locked doors, or barricades with posted signs and flashing lights for isolated areas without enclosures.) i WCGS-Description of Changes to CTS 6.0 10 5/15/97
CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 2 of 8 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAdAY 01 07 Revises Section 6.2.2a. Unit Staff Organization. to reflect No. DCPP is a No. CPSES is a Yes. USAR Chapter Yes. FSAR. LG the non-licensed operator staffing requirements for a multi-unit plant multi-unit plant 13. single unit site. The minimum shift crew composition as described in Table 6.21 has been moved to a licensee controlled document. 01-08 Noves the fire brigade requirements to a licensee No Yes. Hove to FSAR Yes. Hove to USAR. Yes. Hove to FSAR. , LG controlled document. These requirements can be found in LA 75/74 BTP ASB 9.5-1 and their duplication on the ITS is not required. b9 Not used. NA NA NA NA W SF RT 3E5 - 2a_. -} Q 5.7_- l } 01 10 Adds requirement for three auxiliary operators for the two No. Already DCPP Yes No. Wolf Creek is a No. Callaway is a H unit sites with both units shutdown or defueled. requirement. single unit site. single unit site. 01-11 For clarity, a note is added to state that one radiation No. DCPP procedure Yes No. Wolf Creek is No. Callaway is a A protection technician and one chemistry technician can and operational a single unit site. single unit site. fulfill the staffing requirements for both units. requirements differ. 6 01-12 Deletes the Connanche Peak STA qualifications based on use No. Yes No. No. A of RG 1.8. Revision 2. 01 13 Adds new statement to accommodate unexpected absences of No. Already in CTS Yes No. Already in CTS No. Already in CTS A on-duty crew member. 01 14 Deletes the shift supervisors and operating supervisor from No. DCPP procedure No. Not in CTS No. Wolf Creek has Yes A section 6.2 as required to hold a senior reactor operator and operational different license. requiresents requirements differ. 3B -2 g.{QE.2.-t}
^
os es
~
^ nasc.tr as% G[TR So-coS~ }
WCGS-Conversion Comparison Table- CTS 3M.0 $/1587
INSERT 3B-2a 0 5.2-1 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON !COMMANCHE PEAK WOLF CREEK CALLAWAY 01-09 The CTS requirements concerning overtime Yes IYes Yes Yes A being in accordance with the NRC Policy j i Statement is replaced ' by referring to administrative procedures for the control of working hours. INSERT 38-2b Q 5.2-1 j TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 01-15 This change revises the CTS to eliminate Yes Yes Yes Yes-i A the title of " Shift Technical Advisor (STA)." t INSERT 3B-2c TR 5.0-005 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 01-16 A generic title has replaced the plant specific No - Retained Yes No - Retained No - Retained LG utility title for the corporate officer having CTS CTS CTS responsibility for overall plant safety and the plant specific title is moved to the FSAR.
4 Page 5 of 8 CONVERSION COMPARISON TABLE - CURRENT TS 6.0 APPLICABILITY TECH SPEC CHANGE COMANCHE PEAK WOLF CREEK CALLAWAY DIABLO CANYON NUPEER DESCRIPTION , No. Not in CTS. Yes Yes Change the Diesel Fuel Oil Testing Program description for No. Not in CTS. , 02-16 A sampled properties of new fuel oil from "within limits" to i
" analyzed" within 30 days following sampling and addition ;
of the fuel oil to storage tanks. This wording more clearly defines that within 30 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in table 1 of ASTM D975-81 are met. This chan for ITS SR 3.8.3 % ge _zbis jconsistent q g.g., jwith the Bases No. LAR submitted Yes !' nt Pump Flywheel" is being revised consistent Yes No. See Section 02-17 " Reactor C 12/3/%. 1he proposed changes provide an exception to 3/4.4. CN 10-03-LS LS-1 with ; the examination requirements in Regulatory Guide 1.14. Revison 1. " Reactor Coolant Pump Fly f. eel Integrity." godlsedify twSB Yes Yes{QGZ-I} 02-18 Revise the Radioactive Effluent Controls Program dose rate , limits to reflect changes to 10 CFR Part 20.f dra Tes -_Yes A ;
@ LeJWF apK proppfed rtrpweiery- ,_ -fa s.2.-1J No. Addressed in Yes Yes The surveillance interval for verifying that other No. Addressed in ;
02-19 3/4.8 (See CN 01-properties are within limits for ASTM 2D fuel oil is 3/4.8 (See CN 01- " LS-2 60-LS24). changed from "within 30 days" to "within 31 days" after 60 LS24). obtaining a sample. No. Not in CTS. Yes Yes Add the provisions of Specification SR 3.0.2 and SR 3.0.3 No. Not in CTS. 02-20 , A are applicable to the Diesel Fuel Oil Testing program. This change is consistent with TSTF-118. , Yes Yes Yes The method for submitting all reports is revised to be in Yes 03-01 A accordance with 10 CFR 50.4. 01-21 leet.T 3B - 5bp Mc S o-00+ ) i oz-22. msecras-sap-G s.2-1 j A - tuser:T 3s- Sc- }.--[T>c. s.o-co+] 5/15)97 j WCGS-Conversion Comparison Table - C153M.0
r
-INSERT 3B-Sa 0.5.2-1 TECH SPEC CHANGE APPLICABILITY
- NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-22 The Radioactive Effluents Controls Program Yes Yes Yes Yes A is revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3 are applicable to these ;
activities. . INSERT 3B-5b WC 5.0-004 i TECH SPEC CHANGE APPLICABi!ITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-21 Amendment No. 106 for Wolf Creek No No Yes No A incorporated a footnote to allow the volumetric and surface examination of the RCP "D" motor flywheel for the first 10-year ISI interval be delayed for one operating cycle. The examinations are completed during the ninth refueling outage. Since the footnote is a one-time exception and has been satisfied, the footnote is no longer applicable and can be deleted. INSERT 3B-Sc DC 5.0-004 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-23 DCPP Administrative Programs. CTS 6.8.4.d. Yes No No No LG " Backup Method for Determining Subcooling Margin." and 6.8.4.f. " Containment Poar and Turbine Building Cranes." were evaluated for reloaction outside the TS to a licensee-controlled document consiste with 10 CFR 50.36 screening criteria.
._._m.. - _m._... _ _. _______..___.m_ . _ _._.m.-_________________m_ _ _ _ _ _ . _ _ _ . -
9 CONVERSION COMPARISON TABLE - CURRENT TS 6.0 - Page' 8' of 8 ; TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK ICLF CREEK CALLAWAY 03-15 Adds refueling boron concer.tration limits to COLR. Yes Yes No.' Already in CTS. lYes M 03 16 Deletes one of the allowed ECCS evaluation models for CPSES No. Yes No. No. . A Unit 2 which is no longer used. Deletes the methodology section references in the COLR. 03-17 No. References do Yes Yes Yes t A not exist in DCPP L CTS. i 03-18 k Ls c.i Moves cha report st requir PORVs saf for valve ot ton taly if Ofeek pr pr ter p QS.2.-l)
%_ hly . ating R . au sess.y 3 3 - s j 03-19 "The__ "~ t __ed' cha to " i _p .Yes eJa Yse- NA Xee WA !
@ l Hi R ion ea ion. pr i f wnt , -
Yse]ur t I 4 * ~'" I tr _ s di s n sec 1. RG_ St thed . . [ [ 03-20 The use of a continuous guard is provided as an additional Yes Yes Yes No. Maintaining ! LS-3 option for preventing inadvertent entry into high radiation CTS. [ areas that are accessible to individuals. ! I i i [ P i WCGS-Conversion Congperison TnMe- CTS 3M.0 5/1SM
INSERT 3B-8a 0 5.2-1 TECH SPEC-CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 03-18 The CTS requirement to provide Yes Yes Yes Yes LS-5 documentation of all challenges to the PORV's or safety valves is deleted.
1 l 1 , l ) N0 SIGNIFICANT HAZARDS CONSIDERATION (NSHC) l CONTENTS 1 I. Orga ni zat i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 II. Description of NSHC Eval uations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l
- III. Generic No Significant Hazards Considerations 1 l
1 A Aditi ni strati ve Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 1 l \ j R Relocated Technical Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1 l LG - Less Restrictive (Moving Information Out of the Techni cal Speci fi cations) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 M More Restri cti ve Requi rements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 l 1 i IV. Specific No Significant Hazards Considerations LS i LS .1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS 2.................................................................... 16 LS 3.................................................................... 18 LS 4_.................................................................... 20 (LS-f [ t WSOLT 4 Q S . 2. - 1 1 WCGS-NSHC's-CTS 6.0 1 5/1S/97
l l INSERT 4-a [05.2-1) l IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-5 10 CFR 50.92 EVALUATION FOR I TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The CTS requirement to provide documentation of all challenges to the PORV's l or safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694, "THI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 l states: " Assure that any failure of a PORV or safety valve to close will be l reported to the NRC promptly. All challenges to the PORVs or safety valves l should be documented in the annual report." NRC Generic Letter 97-02,
" Revised Contents of the Monthly Operating Report" requests the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does I not specifically identify the need to report challenges to the pressurizer l
safety and relief valves. This change is consistent with NUREG 1431 as modified by TSTF-258. l This proposed TS change has been evaluated and it has been determined that it i involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:
"The Commission may make a final determination. pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a factif ty licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration. if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety."
The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change would not affect the method of operation of plant systems and involves only the deletion of reporting any challenges to the PORVs or
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATION NSHC LS-5 (continued) safety valves. Reporting of challenges to the PORVs or safety valves has not impact on any accident previously evaluated. Therefore, the proposed change would not result in a significant increase in the probability or consequences of a previously evaluated accident.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The assumptions of the accident analyses are unaffected by the proposed change. No new permutations or event initiators are introduced by the deletion of this reporting requirement. Therefore, this proposed change would not create the possibility of a new or different kind of accident.
- 3. Does this change involve a significant reduction in a margin of safety?
The accident analyses are assumed to be initiated from conditions which are consistent with the Technical Specifications Limiting Condition for Operation. The-proposed change does not affect any LCO. Therefore, there is no change in the accident analyses and all relevant event acceptance criteria remain valid. Further, the proposed change has no affect on any actual or regulated failure point which is protected by an event acceptance criterion. Because there is no change in any failure point nor in any event acceptance criteria, there is no reduction in a margin of safety. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS-5" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c), and accordingly, a no significant hazards consideration finding is justified. i I l
_ .. .= - _ - _ _ _ - - . . - ... _- __ .- ... -- - . -. .- -- INDUSTRY TRAVELERS APPLICABLE TO SECTION 5.0 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated NRC approved. TSTF-37, Rev.1 Incorporated 5.6-2 DCPP only. _ TSTF-52 Incorporated 5.5-4 ( qs _ [)Q 5.u -6 \ TSTF-65 @ Incorporated NA MNRC approved { ag eas) Q s.2 -9 4 t* *1 d e s a , , ) TSTF-106, Rev.1 Not Incorporated NA Retain CTS. Incorporated %ee<@lTr F.o-coc.} TSTF-118 5.5-8 __
@l1V' jet 1negrated - ' ,,A A /Petaja CJ}8 l re. c.o -oor. J TSTF-120@ Not Incorporated NA Retain CTS ITn c.oc(- 1 M7 Ipfor60suffeV_ / Sa N ' o 6.2-1]
TSTF-152 Incorporated 5.6-4 %Q 'LTR r-o-co( \ (JFfyAtik ,/Jac6cp(r'ateff / 7 St7C F@ $- -d hp] Incorporated 5.6-5 hipproQy TR 5.0-013 ] WOG-72 Incorporated 5.5-13 I* * \ Incorporated 5.5-14 peppeed Tritteler Incorporated s.s-2, 5.5-1 s.2.3, upacji6 og,2,,, g r.sTF - 2 56 s 2.Il s*@~ *,* 8'#~N
., e 4- .
l t S/158 7
I l Organization 5.2 5.2 Organization 5.2.2 Unit staff (continued) assigned for exh centrol res fra W.i2 ; res;ter i; egr;tini @$d2MiM M thslunft ls?.in MODES 1, 2, 3, or 4. I Uw; unit ;ita with both unit; isthr, er defaled r;;uir; ; t;t;l of thra nen lic;nxd egreter; for tk tw; unit;.] M dB
- b. At in-t ra lianxd "n:ter Ogr;ter (^0) ;bil k pr;,,...t in tk centr;l res Jan fu;l i; in tM r ster. In ;dditis ,
a7mM'E
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. 2. , . . _ _ . ,., . .. .. ,., . .. .. . . - , . . . . . . . . . ,s.
0;nier ",;nter Ogr;ter (L"0; cell b; pr;xat in th; centr;l reem:- i eb. Shift crew composition may be one less than the minimum gggggy I requirement of 10 CFR 50.54(m)(2)(I) and 5.2.2.a and 5.2.2.gf for a period of time not to exceed 2 hours in order to acconnodate unexpected absence of on duty shift crew members , provided immediate action is taken to restore the shift crew I composition to within the minima requirements. dc. AIMMjylduali@ouLthe_HealthiPAysicsjLrgupJsitA11(igDD redtatibn protection procedGris ",sith "hy;i;; TxMicia shall "7 "1 be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position, ed. Administrative procedures 1 hall be developed and implemented to g g gj limit the workina hanes of Urrf74RDwho perform safety
@ related functions (e.g.hicensed Senior;ReactorDrator ^
(SR0s), licensed Reactor:Op~ e ratdr (R0s),; health physic 4Ms El techniciaris, ;ai'id[55 fife?inucleafstationiope' rators, and 1 sit-3 ' key maintenance personnel . y q s.z- 3 Ad;;;t; iift av;r g; 2:11 b; mint;ind witbut rati= k;y u;; cf ;;;rti;;. "a obj;;tiv; 2:11 k to Mv; egr; ting gr caral werk a [" ;r 12] b;r d;y, ;;;in;l 40 Mur nk whil; th; unit i; eg r; ting. llcw;;;r, in th; ;;;r,; th;t unfer;xa pr 21 s ; r;quir; ;ub;t ati;l x~unt; cf ;;;rti n t; k und, er during at;nded gri;d cf duthr, for r;fxling, mjor l l i f (continued) i WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 3 5/1587 l l
l Organization ! 5.2 ! I 1 1 5.2 Organization 5.2.2 Unit Staff (continued) __2_i______ __ __2_._ _,__i __ J 2 ,2 _ _ u _ _ __ _ i- . _ _ _ . . L__2_ .L. Issu u s a bwu suu rbs . vs aveu g v 5 y s us sb rvsussasswu bswit, vuu sa bwissyva un J wu d s .a ha vs l s.1 1_. 2 ._ _ _. . J J 1 2 _ _ _ _ L 11 L. E.11 ._J. swisvwsity yv wsiis~d di su i s ww i v s s v,,w w . l l l 1 A 2 ._ J 2 2 J. . 1 _ L . 1. l -L L- .-- 1LL_J L. .--_L ---- LL__ j &. nus earwsviwuu4 di runs I %s Nb ww g T,s aus a b bwu hv ws vs ruuv g w bssuus L_..__ _A__J Ls . . _1. . J 2 ._ _ _ A..--... A2__. s.e. , ~us J db. usvisb. wmw s uu s i ., .,L,5 2, sA sb us s ~ v w s b s ..~ . n a_ 2 _ J J . 2 J. . 1 _ L_ . 12 __A L_ _.__2ss.; A. ___L ____ AL__ s., e nua a a tw u v s wu%s s .71 Twu 3 %s t TV b ww yUs suu I b bww bw vvv 5 r% ruwws w ba rut 4 i l 1e L_..__ 2.ti u,__..y wn,n L_.._ ___2.J _ _ . _ ____ LL__ n b ,n L...__ 2,._
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nr y gew y y u e, I we e u a vers byWW wwwvb yu s ww a i s twd da ru a u ww uns bI UVs s soww Ias J..____ L. AL. Pn1__t e . ._ _ _ 2 _A _ _ J _ -_1 1 ._ L2_ J__a-__ 2_ uns v us us,w wJ burw L1 auiab .sgsyw p i a l bwa rgsws r bJ vs 33 s d geway a yv rwwe ugg _____J ___ . .J A L ___.__.._J . L_2 ._ 2 A ._ _ A 2 . . . _____J..___ __ L.. L2_L__ ugewvn %sua swW v5 s bs s uyys v v wss uusn i s I a d bs u b a v w ys wwwssus U a , V1 WJ s a 4 ya sw e 1 _ . . _1 _ .s_____----u 2. _____J ___ . 2 AL __A_L12_L_J _.____J..___ iwwwae vi usui ru gwusw s i b , ega uwgevs wggs www av s b e a w e huw s a es swgs ys wwwgsgss ud _J .2LL J._.--a_L2_ _e LL. L__2_ 2__ .___L2.__ LL. J...a_L2__ usvu vu s b u s wwwtssuws u bu b a vs s v3 hesw vgs.a i d u vi yi ussb s 33 ba N uw w a u b e vi s e e__A_i. _L_11 L. 2 1. . J J ,_ AL. ______J..___ _...L AL A 2 Ja . 2 J. . 1 dwwu a be rge b uvvusyuwuua i wvii b y v s d di swa a a ww s s tw a usswws aau byvs yiwwwwuss. l _..__u__ _L_11 L. _ _ . 2 ._ J -__iLi., L.. AL_ r n i _ __i e..___2_i__J__i, wwwe b a vers d u Iw a I ww I wy a wwwww Tvavi I bl i I J ug bsrw Li svieb orwyw a s s ibws wwws i bJ __ L2. J._2____ L. _ _ _ . . _ _ AL A _..____.J .. L...__ L_... __A L.___ vi isiJ ww d a yi Nw bv wa s aus v bi su b Enhwddsvw u Nui a u ess v w I sv k wwws a _ _ _ 2 -- J n .12 _ _ J_.. _L2__ n-__ LL. _L_... _ . 2 J.1 2 ._ _ _ 1. -_L ud d a 33 swws . nwu b a i FE ww v s u w a vs s 4 I vvus bstw uwwvw yu I ww a u u zwd id a rV b _ . .A L _ _ 2 _ _ J me u~. -. _ . s,2.s J QuseAT S.6-4_) g s s2-1 3 s1 - ! ITIEamo of ove work unit staff members orming saf relat unction all be limi and co led in l l corda with t Policy St nt on ing ho (Generic j Lett _82 12). _ n_ _ _ _ u _ _ _ u______ ,__2_a__, ~___u___ u.____- j e pi #E Vywi u b a VI 3.5 5 rui ruyw s __ ndJidbugIb vywI u%gvisd I r%st Buys s VI
~ ~ - -
SupeMntendent1 Operations or. Manager _0peratfons;shall hold an SR0 license. t j (continued) WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 4 S/15/97 r
i I
~NSERT~5.0-4a
. 0 5.2-1 The. controls shall include ' guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime. Any
' deviation from the above guidelines shall be authorized in advance by the E Plant Manager ) or the.E Plant Manager's 4 designee, in accordance with i approved administrative procedures, and with documentation of the basis for granting the deviation. Routine deviation from the working hour guidelines 1 shall not be authorized.
l l l 4 I
i l Organization 5.2 5.2 Organization
. An inMvielu.d
}4 s.2.-t {
gf. MeeI51swrf d_NM shajlprovide advisory technical
~ n ".: . y support to thESTMC.sd[gpssirJ8SDTn the areas of thermal i u g,o. m j hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In eddition, th; STA ; hell l u
hpigsition"shall be manned ~Lin: MODES ~112; 1 gfj4l"unless E
- Gb m theMt3heJnd.i.Vidual;withj a;SenioChratotiyicense~; meets 3"*9eD the qualifications specified by the Commission Policy Statement on
,,, , 4 Engineering Expertise on Shift.
I i 1 l l [ WCGS-Mark-up ofNUREG-1431-ITS S.0 5.0 5 S/15m l
1 Unit Staff Qualifications 5.3 l 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications l ";vis;r's t';tc. tiinir.u;; qualifications for ...bcrs of th; unit st;ff sh:11 b; i sp;;ified by u ; cf cr. ov r:11 qualifistion stetse.-,t ref;rcacing er, ".'O Stud;rd x;;pt;ble to th; !"',C steff or by spuifying individu;l p;sition qualifiations. l 0:ncrolly, th first athed is pr;ferable, however, th; xand athed is ; dept;bic i to th;;; unit ,___,
-tsffs requirir.; puiel qu;1ification statacats beaux of ur,ique 1
.. ,u..,__
..u.....u.
...m....___ u. . -
l 5.3.1 Each member of the unit staff shall meet or exceed the minimum
$513?It#s qualifications of ANSI /ANS 3.1-1978 with the..following:.ex_ceptions;
~ ~"
[",cgul tery Cuid: 1."_<, "_,;;i-ica 2, 1^07 cr r.or; r;;;nt revisions er
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.. _ _ _ _ A . L 1 _ A_ tana _m.s uwusywww.w wv rwu a s u . iJ. si 5;81U1 t!feersusMwr~aturwit!"sentutweysturnstmTraueretrar gyg2Jyj exceedMaunlificationsidffANSpANS T1 1981'agendo,r. sed byJegulatory31de;1;8;1tevisibnTend'J0 EFRTart255.
___ ;;-tiangeF cnmars O Lwc s.o-m3 \ 5.3.1.2 T jhesition,oJ[37.MQRadiation. Protection shal] z meet"orfexceed.ithii qualificatio~nsiioERAgulatorEGuide~1i8 Septembeir 1975 for a lladiationProtedtT6nianagerl 5.3.132 The_ position of Manager 0perationsjshal_lihol_dior;haye previously held a senior reactor operator 71 cense.for.'a Sim.ilar_pgitJPWR). S .L 2- Ec,e th pr ou. clr to C.FR 55.24, o_ hCmd. Sexuh_ bdn Cpuahn- (5 D) M a licw' o A re.x b r e,y2.br C R.c) cig c. th os,a. vA ch vi ca.nls, u.Mo, A a etcutiuw te we.i k' h$' h egd1*" =4 4 T5 S.3. ), .pa< few m h b c.t tous d.a.s cd \24 d - N IO CC4 SD. 514-(*') .
-~
l l l l l l l WCGS-Mark-up ofNUREG-1431-1TS 5.0 5.0 6 S/1S/97 l 1 l l
l , 1 Programs and Manuals 5.5 5.5 Programs and Manuals and setpoint determination in accordance with the methodology in the 00CH: I
- b. Limitations on the concentrations of radioactive material
, released in liquid effluents to unrestricted areas, conforming l to 10itimesjthelconcentrationNalueshin 10 0m 20, Appendix B, l$S$53$@ Table 2. Column 2Tto'101CFR;20!100120 2402:
- c. Monitoring, sampling, and analysis of radioactive' liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the OOCH:
- d. Limitations on the annual and quarterly doses or dose commitment i to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I:
- e. Determination of ctmulative ;r.d pr;jated dose contributions from radioactive effluents for the current calendar quarter and $$5%$
current calendar year in accordance with the methodology and -4 parameters in the 00CH. t 1 c ;t e n ry 31 d;y;. Iptiersipat1M L pjro egedidoselcogttbutionsifgregoa~cMEsfflunnts3: i !I@n#hlceNthWheth6doloMnitMMat3tqVUYory.131 daysi
- f. Limitations on the functional capability and use of the liquid and l gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose comitment, conformina
( 6_ h sas7 to 10 CFR 50 Ap@pendix I:
- g. Limitations on the dose raYe\resulting fromradioactivematerial@S'II released in gaseous effluents *to areas'beyond the site boundary Cconfir@@t; the d;;; ;;;ociq"d with 10_0m 20. ?+Er. dix 0. _
isfS!SE11d T;ble ?. 0;1= 1, @Aff#e . sWin be hecoMmu.3 c_m %. f.no@ : J l 1; ForEnobleJase_s; Les(then orieuaPtB aidosejrateS500' mrem /yr to'the whole body a_nd C esruu rorpansa do_se - rat _eM em/yr to the.(skin;3 nd
- 2. For_Iodinejl31,1 f or Iodine 133,.for;triti.ta,;and;for all WKFiSMM (continued)
WCGS-Mark-up ofNUREG-1431-ITS S.0 5.0 10 S/1S/97
Programs and Manuals 5.5 l 5.5 Programs and Manuals , ratf[qnt@Tae5IQparticuTee@mjdtE_hiTfSTyes gts# term.~,5 i dg g 3p m N Ja'_dosg rate _ _ aren/,g2tejany }gg,g_g oten _
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I:
I. Limitations on the annual and quarterly doses to 'a member of the public from iodine 131, iodine 133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, confoming o 10 CFR 50. Ap:endix I: _and g*g g pe.sina.% owe. koaw3g
- j. Limitati s on the annual dose or oose commitment to any member of%.2.-li the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
-r 5.5.5 Comoonent Cvelic or Transient Limit This program provides controls to track the UFSAR, Section 3.9(N),
cyclic and transient occurrences to ensure that components are gg:g~ maintained within the design limits.
- k. The ov6bnu of s 3.o. 2. >eA. se. a.c . 3 M, Q 3T i c.airL. Io the. Rach*o achet. Ef-flue 4 ContreRs 1-los.t.-t j i
% y m s a n :. 1 % a & y 9 I
(continued) WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 11 S/1S/97
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Monthly Oneratina Reoorts Routine reports of operating statistics and shutdown experienc .s ~. o s ety y esy n submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. s ,2.4 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established' prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
"; indivii;l :p;;ification; th;t '" ;;; cer; ep;r; ting li;it; mu;t b; r;f;r;rced ter;.
ggg Is Specification 3.1.3:11ModeratorTempera_ture Gatffdsimt20frgh 21 SigRf c~at]on ~33.5j; Shutdownliank Jnserti6rilialts; 3I SpeciffcatYon;3;T.6FControT"Benk" Ins ~ertT CTafts 4; Specification 3.2.3: Axial R.uxRifyy, . t,.
- 5. Specification 3.2.11~HeatJg,HotjChanneljfactoriffZ), j 6.. Specification 3.2.2: Nuclear.Inthal_py, Rise; Hot ~ Channel FactoE(F",,) .
- 7. Specification. 3.9.1:iBoron Concentration?;and 8._ SHL@0WN~ MARGIN for. Specification l3.1'.1 and 3.1.4, 3,1.5 336Eand]I3.8.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
Identify the Tepic;l R; pert';) by ag.b;r. titi;. it;. ;nd l00 gg teff ;pprove.1 icu. cat. er identify tre :teff 0;f;ty Cvela;tica 0; pert fer ; pier,t ;p;;ific ;thodeled by l00 letter c.ad it;. (continued) WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 31 S/15M7
[ 5.0 @ 9. 2.- t High Radiation Area (5.T-] l 5. ADMINISTRATIVE CONTROLS I gh Radiation Area 5.7 gypgg 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 0 CFR 20, in which the intensity of radiation is > 100 mrem /hr but s 000 mrem /hr at;30 cm (12 in'.), shall be barricaded and ig517]ib co picuously posted as a high radiation area and entrance there to shal be controlled by requiring issuance of a Radiation Work Permi (RWP). Individuals qualified in radiation protection proced es (e.g., Health Physics; Technicians) or personnel _ y g
,q^
continuo sly escorted by such individuals may be exempt fr the RWP j issuance uirement during the performance of their ass duties l in high ra 'ation areas with exposure rates s 1000 mr hr, provided they are ot rwise following plant radiation protec on procedures for entry into m,5 high radiation areas. Any individual group of individuals penni ed to enter such areas shall be provid with or accompanied by o or more of the following:
- a. A radiation nitoring device at continuously indicates the radiation dose rate in the a a.
- b. A radiation moni ring vice that continuously integrates the radiation dose rat i the area and alarms when a preset integrated dose is eived. Entry into such areas with this monitoring devic made after the dose rate levels in the i area have been stabli and personnel are aware of them,
- c. An indivi al qualified i radiation protection procedures with a radiati dose rate monitor g device, who is responsible for provi ng positive control o r the activities within the area a shall perform periodic rad tion surveillance at the equency specified by th; ";di; ica "rct ction ";n;;c health 4B;PS W
physics supervision in the RWP. 5.7.2 addition to the requirements of Specific ion 5.7.1, areas with .. radiation levels a 1000 mrem /hr at:30;cm (12_ _.) shall be provided * !j5Rf1% with locked or continuously guarded doors to p . vent uneutheriz;d g*gg inadvertent entry and the keys shall be maintai under the administrative control of the Shift Supervisor 1 / rvising Operator ygpsgp (continued) WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 37 S/158 7
. __ . _ _ _ _ _ . __ .~_ __ _ __ _ . _ . _ _ - _ _ _ _
High Radiation Area 5.7 l 5.7HighRadiationArea j q s.t.- I } 5.7.2 (con ued) , I i Shift . ;;r. on duty or health physics su ision. Doors shall remain lo except during periods of ess by personnel under an approved RWP at shall specify the e rate levels in the immediate ! work areas and maximum allowa stay times for individuals in those areas. In 1 of the s time specification of the RWP, direct ! or remote (such as c sed c cuit TV cameras) continuous surveillance may be made by personne ualified in radiation protection procedures to provide positive e s e control over the activities being l l performed within t area.
- l
\ l l 5.7.3 For individua igh radiation a as with radiation levels of
> 1000 mr r atJ0ica (12 in.), ccessible to personnel, that are gggggg located thin large areas such as actor containment, where no "
enclo re exists for purposes of lock g, or that cannot be co nuously guarded, and where no enc sure can be reasonably nstructed around the individual area, at individual area shall be l barricaded and conspicuously posted, and a ashing light shall be l activated as a warning clevice. l l l l l I l l l l WCGS-Mark-up ofNUREG-161-ITS S.0 5.0-38 S/ISA 7
I liigh Radiation Area 5.7 5.0 AD!!INISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20. the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20: 5.7.1 Hiah Radiation Areas with Dose Rates Not Exceedina 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: l
- a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel l or equipment. I
- b. Access to, and activities in, each such area shall be controlled l by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate !
work area (s) and other appropriate radiation protection equipment and measures. 1
- c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise
)
following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d. Each individual or group entering such an area shall possess: j
- 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or
- 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
- 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or l
(continued) l
l High Radiation Area 5.7 I 5.7 High Area Radiation Area 5.7.1 Hiah Radiation Areas with Dose Rates Not Exceedina 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by l the Radiation: (continued)
- 4. A self reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and.
l (1) Be under the surveillance, as specified in the RWP l or equivalent, while in the area, of an individual qualified d.a radiation protection procedures, l equipped with a radiation monitoring device that l continuously displays radiation dose rates in the l area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in l radiation protection procedures, responsible for
- controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
- e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.
5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by { the Radiation. but less than 500 rads / hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation:
- a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition: 4 ygwr.s o-ca3]
l 1. All such door and gate keys shall be-maintained under the_ administrative control of theMacoerVWo3gpcat>on ~ py.otecperryafGkJe#, or his or her designee. hesi% phgdcs supavisoD (continued)
l - incest 5.0-31 conh High Radiation Area l
'5.7 High Area Radiation Area
,5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rea/ hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Meter from the ;
Radiation Source or from any Surface Penetrated by the Radiation: l l (continued)
- 2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit. I
- b. Access to, and activities in, each such area shall be controlled i by means of an RWP or equivalent that includes specification of i l radiation dose rates in the immediate work area (s) and other i- appropriate radiation protection equipment and measures.
l-
- c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d. Each individual or group entering such an area shall possess:
- 1. A radiation monitoring device that continuously integrates
! the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an i appropriate alarm setpoint, or
- 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote I
receiver monitored by radiation protection personnel l responsible for controlling personnel radiation exposure l within the area with the means to communicate with and ; control every individual in the area, or ! l i
- 3. A self reading dosimeter (e.g., pocket ionization chamber '
'or electronic dosimeter) and.
l (1) Be under the surveillance, as specified in the RWP
- or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (continued) i
- , ~ - - -- ~ , , - , , , . . _ .
he.a.T S.ihl coEtO High Radiation Area 5.7 5.7 High Area Radiation Area 5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Heter from the Radiation Source or from any Surface Penetrated by the Radiat h (continued) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area, or
- 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.
. Except for individuals qualified in radiation protection I
e.scorted b Sd procedureJ, entry into such areas shall be made only after dose t dd nidu.816 rates in the area have been determined and entry personnel are M s.2-l] knowledgeable of them.
- f. Such individual areas that are within a larger area M3 (co3t.PdYlyids 3.Mgh Pid5[>ETonMrJD where no enclosure exists 7,4 3
succhu PWIR for the purpose of locking and where no enclosure can reasonably ggj,,q be constructed around the individual area need not be controlled co d dnvnent' by a locked door or gate nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device. (continued'-
DIFFERENCES FROM NUREG 1431 ! Section 5.0 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431 Revision 1. to make them plant specific or to , incorporate generic changes resulting from the Industry /NRC generic change process. The change nebers are referenced directly from the NUREG 1431 mark ups. For L Enclosures 3A. 38, 4, 6A, and 6B. text in brackets "[ ]" indicates the information l 1s' plant specific and is not comon to all the Joint Licensing Subcomittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific l information in that location. CHANGE j Nl#EER JUSTIFICATION 5.1 1 Revises Section 5.1.1 to maintain current TS. The Plant Manager does not currently approve prior to implementation each proposed test. experiment or modification to systems or equipment that affect nuclear safety. The design process includes design verification by qualified persons to assure that the design is adequate and meets specified design input. Design / configuration changes and test procedures that require a written evaluation under 10 CFR 50.59 are reviewed by the Plant Safety Review Comittee prior to implementation. There are adequate administrative controls for review of proposed tests, experiments or modifications so that review by the Plant Manager is not necessary. 5.1 2 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 5.2 1 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B).
.5.2 2 This change deletes Section' 5.2.2.b since the requirement for the presence of a reactor operator (RO) or a senior reactor operator (SRO) in the control room is adequately controlled by 10 CFR 50.54(m)(2)(iii) and 50.54(k). The ITS 5.2.2.b requirement that is being deleted will be met through compliance with these regulations and is not r uired in the TS. This change is consistent with traveler TSTF 25 'Q S t-l]
5.2 3 Qletaned. i)SEUIL hA - l A QS.2-I} l 5.2 4 Section 5.2.2.a describes the unit staff requirements for non-licensed operator staffing for multi unit sites. This change reflects plant specific requirements for a single unit site and is consistent with the current TS. l 5.2 5 Not used. I WCGS-Di(ferencesfrom NUREG-1431 - ITS 5.0 1 S/15A7 l l
. _ - _ _ _ . . . _ _ _ _ _ _ . _ __._ -._._ _ _. _ . _m _ - _ _ _ _ _ _ _. _ __.
INSERT 6A-la 0 5.2-1 5.2-3 ITS Section 5.2.2d (ISTS 5.2.2e) is revised from specific working hour limits to administrative procedures to control working hours. The proposed changes will provide reasonable assurance that safe plant operations will not be jeopardized by impaired performance caused by excessive working hours. Specific working hour limitations are not otherwise required to be in the technical specifications under 10 CFR 50.36(c)(5). Specific controls for working hours of reactor plant staff are described in procedures that require a deliberate decision making process to minimize the potential for impaired personnel performance, and that established procedure control processes will provide sufficient controls for changes to that precedure. These changes are consistent with the
- recommendation in the April 9,1997 letter from C. Grimes to J
! Davis. Additionally, the ISTS statement " Controls shall be l included in the procedures such that individual overtime shall be
- reviewed monthly by the [ Plant Superintendent] or his designee to ensure that excessive hours have not been assigned." is being j deleted. There is no guidance in Generic Letter 82-12 that
! discusses these additional controls. The additional requirement to have the Plant Superintendent (or his designee) review individual overtime on a monthly basis is unnecessary since l sufficient administrative controls and policies exist, as well as ! the role of the individuals supervisors in supervising personnel l prevent excessive or abuse of overtime. These changes are < consistent with TSTF-258. L i I l l-I i
(3,s..e ~ Not ape _nc2A b WCM 5*c Gmesn Comymww TECCwbun ta)). loc s.o-oo2.)
.[' @ -q t4.hplicaxa.buccr5,. see emersim compmww TAA4. (E.wdomuAsh1w.s.o.cos)
CHANGE NUMBER JUSTIFICATION 5.2 6 Thispshgere s SecJt m Sp2f to.describedcur.penrTTS] .5er 'The Shfft T_ _al Adv.iaaf (STAT. lkis.e*T GA -h j e s.2.-i T 5.2-7 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 5.3-1 This change revises Section 5.3.1 to be consistent with current TS regarding plant staff qualifications and training. Q-2 ~ INSES.T 6 ATTE)---l Q S.2.- t j These changes revise Section 5.5.4, " Radioactive Effluent Controls 5.5-1 Program." to reflect new 10 CFR Part 20 requirementsJ {, ,, g , } rav 13Meinggnerfeti topilecra revision ter G1 1 sistenf witV10 CFTPart 20. lusee.T 64 -2c. l l 5.5 2 This change revises Section 5.5.3, " Post Accident Sampling." to ensure I the capability to obtain and analyze radioactive " iodines" in lieu of l
" gases." This change is consistent with the current TS and plant practices.
5.5 3 This change revises Section 5.5.8, " Inservice Testing Program," to l delete " including applicable supports." This change is consistent with the current TS. 5.5 4 The Containment Leakage Rate Testing Program is added to the improved Technical Specifications (ITS) consistent with the current TS. The Containment Leakage Rate Testing Program is consistent with traveler TSTF 52. 5.5 5 This change revises Section 5.5.13. " Diesel Fuel Oil Testing Program," to be consistent with the current TS. The details of the method l applied to this test are discussed in the associated SR 3.8.3.3 Bases. [To maintain consistency with the Bases for 3.8.3.3, specific changes to the program description are for sampled properties of new fuel oil from "within limits" to " analyzed' within 31 days follaing sampling and addition of the fuel oil to storage tanks. This wording more ' l clearly defines that within 31 days following initial new fuel oil .! l_ sample, the fuel oil is analyzed to establish that the other priorities are met.] l 5.5 6 Not applicable to WCGS. See Conversion Comparison Table (Enclosure l 68). 5.5 7 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 5.5 8 A sentence is added to Section 5.5.9 ("The provisions of SR 3.0.2 are applicable to the Steam Generator Tube Surveillance Program test frequencies.") and Section 5.5.13 ("The provisions of SR 3.0.2 and SR WCGS-lhfferencesfrom NUREG-1431 -ITS 5.0 2 S/158 7
l INSERT 6A-2a 0 5.2-1 l ! This change revises ITS Section 5.2.2.f (ISTS Section 5.2.2.g) to describe the current [TS] and to eliminate the title of " Shift Technical Advisor (STA)." , STAS are not used at all plants (the function may be fulfilled by one of the l other on-shift individuals). This Section is revised so that it does not i imply that the STA and the Shif t Supervisor must be different individuals. l Option 1 of the Commission Policy Statement on Engineering Expertise on Shif t l 1s satisfied by assigning an individual with specified educational qualifications to each operating crew as one of the SR0s (preferably the Shift Supervisor) required by 10 CFR 50.54(m)(2)(1) to provide the technical expertise on shift. However, the ISTS 5.2.2.g wording of, "the STA shall provide ... support to the Shift Supervisor...." is considered to be easily misinterpreted to require separate individuals. Therefore, the wording is revised so that the STA function may be provided by either a separate individual or the individual who also fulfills another role in the shift
' command structure. This change is consistent with TSTF-258.
INSERT 6A-2b Q 5.2-1 5.3-2 New paragraph 5.3.2 is added to ensure that there is not misunderstanding when complying with 10 CFR 55.4 requirements. The Definitions in 10 CFR 55.4 state: " Actively performing the functions of an operator or senior operator means that an individual has a position on the shift crew that requires the individual to be licensed as defined in the facility's technical specifications, and that ...." Placing this paragraph in the ITS meets the 10 CFR 55.4 requirement for defining in the facility's technical specifications the function performed by licensed individuals per 10 CFR 50.54(m). Adding this paragraph is consistent with the recommendations in the April 9, 1997 letter j from C. Grimes to J. Davis and TSTF-258. l INSERT 6A-2c 0 5.2-1 After issuance of Generic Letter 89-01, 10 CFR 20 was updated. The NRC issued a draft Generic Letter, 93-XX, on proposed changes to STS NUREGS based on the new 10 CFR 20. The proposed changes are consistent with the draf t generic letter, the April 9,1997 letter from C. Grimes to J. Davis (with some exceptions) and traveler TSTF-258. The proposed changes maintain the same overall level of effluent control while retaining the operational flexibility that exists with current TS under the previous 10 CFR 20. These changes are l intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements. l
i s.s -e U frl'o A -b m. A 6 _44w 6p mMa&(enthau]u.ec,5,o.coi} HANGE ' NutEER 4,s.Tg ~ tusEnr eg.4Mcp s.s J j dl$TIFICATION 5.5 15 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). Q.s-ic._ tT4se =T o 4 - h ) o s.2-i 1 l 5.6 1 Not applicable to WCGS. See Conversion Comparison Table (Enclosure s.si61 ~ iN d. crc.A- @] I 5.6 2 This change deletes the Emergency Diesel Generator Report to reflect I the recomendations of GL 94 01 " Removal of Accelerated Testing and l Special Reporting Requirements for Emergency Diesel Generators." dated May 31, 1994. ! 5.6 3 This change revises the report date in Section 5.6.2. " Annual Radiological Environmental Operating Report." to be consistent with l current TS . l 5.6 4 This change revises Sections 5.6.1 and 5.6.3 " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report." i respectively, per NRC letter dated July 28, 1995. " Changes to Technical l Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From l Christopher I. Grimes to Owners Groups Chairs). This change is l consistent with traveler TSTF 152. 5.6 5 lift settings are referenced in the PTLR section
~
Cs. c. -r, iusser (,N4 bN cas.: i1 5.7 1 (This~ ang evise High ja a to rpo c co st with 0C 0. 1]. pecif' ally. 1stan s fr the $ 5. % l tadia on so ce are t . Ius.EM(oA -4 c. j
~
5.7 2 FTh ha r ses " naut riz t "in vert t" nteH on a se ref ct e NR s iti n a sta ed 'n / Re ato Guig8.38 Sec n1 reg din phy cal arr er for/High adia on Areds. T si cons tent ith av er T F- 67. I I 5.7 3 (Not applicable to WCGS. l usEc1 4.A- 4 d See Conversion Comparison Table (Enclosure 6B). 6A- f e-M o s .2- d I WCGS-Differencesfrom NUREG-1431 - ITS !.0 4 S/158 7
.. - ._ _ _ _ _ ~. __ _.. _ ._.._ __ _ _ _ . _ . _ _ . __ _ _ . . _ _ _
INSERT 6A-4a 0 5.2-1 1 5.5-16 The Radioactive Effluents Controls Program is revised to include l l clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. These statements of applicability clarify the allowance for surveillance frequency extensions and allowance to perform missed surveillances. Generic Letter 89-01, " Implementation of I Programmatic Controls for Radiolog1 cal Effluent Technical Specifications and the Relocation of Details of RETS to the i Offsite Dose Calculation Manual or the Process Control Program" l allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents i Control Program in the Administrative Controls Section of the ' Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 89-
- 01. Since this change adopts previous CTS requirements, it is l
considered a change of presentation method only. This change is consistent with TSTF-258. INSERT 6A-4b 0 5.2-1 5.6-6 The ITS requirement to provide documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694. "TMI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states:
" Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or .
safety valves should be documented in the annual report." NRC l Generic tetter 97-02. " Revised Contents of the Monthly Operating ! Report" requests the submittal of less information in the I monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic i letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent with TSTF-258.
. INSERT 6A-4c 0 5.2-1 l Section 5.7 is revised in accordance with 10 CFR 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. These changes are consistent with the draft Generic Letter (93-XX) on proposed changes to STS NUREGs based on the new 10 CFR 20 and the letter f rom C. Grimes, NRC, to l
J. Davis. NEI dated April 9,1997. This change is consistent with TSTF-258 and encompasses the NRC comments on 6/11/98. Additional technical changes made to Section 5.7 are identified and justified.
INSERT-6A-4d 0 5.2-1 ITS 5.7.2.e is revised consistent with CTS 6.12 that allows any individual or group of individuals to enter a high-high radiation area (dose rates greater than 1.0 rem / hour at 30 cm) accompanied by an individual qualified in radiation protection procedures with a radiation dose rate monitoring device. The qualified individual is responsible for providing positive control and , shall perform periodic radiation surveillances at the frequency specified in ) ! the RWP. The CTS requirements allow the qualified individual to enter a ! locked high radiation area with plant workers without first having to enter I the area to determine dose rates and then exit the area to provide dose rate I information to the plant workers and then reenter the area. This flexibility is in keeping with the "As Low As Reasonably Achievable" principle while maintaining appropriate radiation worker practices. INSERT 6A-4e 0 5.2-1 ITS 5.7.2.f is revised consistent with CTS 6.12 to delete the phrase "that is controlled as a high radiation area". The proposed change would preclude having to post an area around the high-high radiation area as a high radiation area when the area may not meet the definition of a high radiation area. 1 INSERT 6A-4f CA 5.0-003 ITS Section 5.5.10 is being revised consistent with CTS 8.8.4.c. The proposed change deletes the phrase "and low pressure turbine disc stress corrosion cracking" from the ITS to be consistent with the practices of the CTS which do not.have this requirement for the Secondary Water Chemistry Program. INSERT 6A-4g 0 5.5-7 5.5-19 CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety related ventilation system filter adsorber units include the requirement to measure flow rates within specified values, while imposing an artificial differential pressure. during system operation, when tested in accordance with ANSI N510-1980. This flow rate testing is to be performed at least once per 18 months. af ter any structural maintenance on the HEPA filter or charcoal adsorber housings, or following painting fire, or chemical release in any ventilation zone communicating with the system.
. Therefore the CTS surveillance requirements are incorporated into j the ITS.
l I i
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 1 of 5 SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLhlAY 5.1-1 Revises Section 5.1.1 to maintain WCNOC current technical No No Yes No specifications (TS). The Plant Manager does not currently approve prior to implementation each proposed test. experiment or modification to systems or equipment that affect nuclear safety. 5.12 Revises Section 5.1.1 to maintain Callaway CTS that the No No No Yes Plant Manager approves prior to inplementation each proposed test. experiment or modification to systems or equipment that affect nuclear safety and are not addressed in the FSAR or TS. 5.2-1 Revises Section 5.2.2.a to reflect the CTS. This change Yes Yes No. Wolf Creek is No. Callaway is a clarifies the application of the unit staff provisions to a single unit site. single unit site. both units. 5.2-2 The requirement for the presence of a RO or a SRO in the Yes Yes Yes Yes control room may be deleted from the ITS since this requirement is adequately controlled by 10 CFR 50.54(m)(2)(111). - W jW (5.2-3 NoCuted._ i WER.T (al!! - l m., _
# . )NA QS.1-1 l 5.2-4 Section 5.2.2.a describes the unit staff requirements for No. DCPP is a No. CPSES is a Yes Yes non-licensed operator staffing for multi-unit sites. This multi-unit plant. multi unit site.
change reflects plant specific requirements for a single unit site and is consistent with the current TS. NA NA NA NA 5.2-5 Not used. 5.2-6 Yes Yes. LA 50/36 Yes Yes [~ Revises section 5.2Jf to describe the currentTS moved text to FSAR j (-swr.
@i c2R.owAAo ellmhts-Mvisie (STAT . 4ba. Sila. wf SdnSection 13.1 which permits on-shift SRO to fill STA QS.2.-l } position.
WCGS-Conversion Comparison Table-ITS5.0 $/15i97
=-
INSERT 6B-la 0 5.2-1 i.1 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON- COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION I Yes. Yes Yes 5.2-3 ITS Section 5.2.2d (ISTS 5.2.2e) is revised Yes from specific working hour. limits to ! administrative procedures to control ' r working hours. 3 i 4 I
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 2 of 5 s,2 - e ine 66 m so-co 2. ] SECTION 5.0 g.2 A ins <t 6s - 2c. ra s.o - cos] DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.2-7 Revises 5.2.2c to add note that a "; ingle Radiation No. Not current Yes No. Wolf Creek is No. Callaway is a Protection Technician and a single rhemistry Technician procedure or a single unit site. single unit site. may fulfill the requirments for both units. operational
- requirement.
5.3-1 Revises Section 5.3.1 to be consistent with current TS Yes. LA 43/42. Yes Yes Yes regarding plant staff qualifications and training. 5.5-1 Revises Section 5.5.4." Radioactive Effluent Controls Yes Yes Yes Yes Program." to reflect new 10 CFR Part 20 requirements. NR / st i. w_y - Q '5."2_- I } Revises Section 5.5.3. " Post Accident Sampling." to ensure Yes Yes Yes Yes 5.5-2 the capability to obtain and analyze radioactive " iodines" i in lieu of " gases." This change is consistent with the current TS and plant practices. Revises Section 5.5.8. " Inservice Testing Program," to Yes Yes Yes Yes 5.5-3 delete " including applicable supports." This change is consistent with the current TS. 5.5 4 The Containment Leakage Rate Testing Program is added to Yes. LA 110/109 Yes. LAR %-002. Yes Yes the ISTS consistent with the current TS. The Containment Leakage Rate Testing Program is consistent with traveler TSTF-52. Revises Section 5.5.13. " Diesel Fuel Oil Testing Program." Yes Yes Yes Yes 5.5 5 to be consistent with current TS. The details of the method applied to this test are discussed in the associated SR 3.8.3.3 8ases. Additional programs are added to the ITS (other than j kkre dkhed)Yes No. No additional No. No additional 5.5-6 Containment Leakage Rate Testing Porgram discussed in CN \ preym programs added. programs added.
~ - 4AApL-- -l DC. SD -lb+1 5.5-4).
{3-2}} msiiirr ss-2.ya_[ o s.2.- I } WCGS-Conversion Conparison Table-ITS5.0 5/15R7
/ ,
7 INSERT 6B-2a .0 5.2-1 . TECH SPEC CHANGE l_ APPLICABILITY NUMBER DESCRIPTION ?iABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 5.3-2 New paragraph 5.3.2 is added to ensure that Yes Yes Yes Yes there is not misunderstanding when complying with 10 CFR 55.4 requirements. i INSERT 6B-2b DC 5.0-002 j TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY , 5.2-8 Revises Section 5.2.2 and 5.3.1 to reflect Yes No No No , License Amendment 128/126 dated 6/11/98 which changed requirements for the DCPP Operations Director. INSERT 6B-2c TR 5.0-005 i TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY A generic title has replaced the plant specific No Yes No No l 5.2-9 ' i utility title for the corporate officer having responsibility for overall plant safety. __ _ _ _ _ _ ___ _ _ _ ___._____________________________________J
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 4 or s SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION No No No 5.5-12 The referenced frequencis for the tests listed in the Yes Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96 09) Yes Yes Yes 5.5-13 Revises Radioactive Effluent Controls Program dose Yes projections to meet original intent of TS prior to implementation of GL 89-01. (WOG 72) Yes Yes Yes 5.5 14 Section 5.5.7 is being revsted consistent with traveler Yes _ [and [aj $ctEng IMtJe(uestdubqiited December e %gQ4 6 [o?, -g3,g.g (P / . The proposed changes to Section 5.5.7 provide an exception to the examination requirements in Regulatory gg yq - ~ gqq l Guide 1.14. Revision 1. Reactor Coolant Pump Flywheel Integrity." No No No Yes 5.5-15 This change provides a time interval of within 31 days l after removal in which a laboratory test of a sample obtained from the charcoal adsorber must be tested. This l change is consistent with Callaway CTS. Yes. LAR 94-14 No. Wolf Creek CTS No. Callaway CTS 5.6 1 Revises Section 5.6.4. Monthly Operating Report." to No. DCPP CTS consistent with consistent with consistent with reflect a revised submittal date. NUREG-1431. NUREG-1431, NUREG-1431. Yes No. Not in CTS No. Not in CTS. No. Not in CTS. 5.6-2 Deletes the EDG Report to reflect the reconenendations of
~
equir f D -O:? 5.6-2_ } s dat ay 31, TSTF- 3 7, Rev i , , Yes. Consistent Yes. See LA 42/28. Yes Yes 5.6-3 Revises report dates in ITS 5.6.2. " Annual Radiological Environmental Operatir.g Report" to be consistent with with CTS and LA current TS. 78/77. thAsat:T (A -khj Q S.5-7 l (S.p l(n NSE5tT 68-4 A. [ o S.z. t } 5.S-let tiahtt 64 - 4bM CA S.O-003 {
' = " a -4 0 -1 oc so e sl G.s-i6tVrce r%...mian $/15/97 re unarisnn TnMs. - ITV % 0
' INSERT 6B-4a 0 5.2-1 ,
TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION The Radioactive Effluents Controls Program Yes Yes Yes. Yes 5.5-16 i is revised to include clarification ; statements dencting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. < INSERT 6B-4b CA 5.0-003 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY , QMBER DESCRIPTION Yes Yes 5.5-17 This change deletes the phrase "and low pressure Yes No ; turbine disc stress corrosion cracking" from ITS , 5.5.10 to make the program consistent with CTS 6.8.4.c. INSERT 6B-4c DC 5.0-003 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON. COMMANCHE PEAK WOLF CREEK CALLAWAY l NUMBER DESCRIPTION Yes No No No 5.5-18 Revises DCPP Sections 5.5.9 and 5.6.10 to reflect License Amendment 124/122. dated March
- 12. 1998, which allows implementation of steam generator tube voltage based on repair criteria for ODSCC indications at tube to tube support plate intersections.
I _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ ____._.__ _. _ _ _ _ _ _ _ . . _ _ _ _ _ ~ _______..____I
4
. INSERT 6B-4d 0 5.5-7 TECH SPEC CHANGE APPLICABILITY-DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION No No Yes No 5.5-19 Wolf Creek CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety-related ventilation system filter adsorber units include ,
the requirement to measure flow rates within specified values, while imposing an artificial differential pressure, during system operation. - when tested in accordance with ANSI M510-1980. ; The CTS surveillance requirements-are ; incorporated into the ITS. ; i
?
I t I f' r E e h
- f
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECflON 5.0 Page 5 of 5 SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY NLM ER DESCRIPTION DIABLO CANYON CONANCHE PEAK WOLF CREEK CALUMAY l 5.6-4 Revises Sections 5.6.1 and 5.6.3 " Occupational Radiation Yes Yes Yes Yes Exposure Report" and Radioactive Effluent Release Report." i respectively, per NRC letter dated July 28. 1995. " Changes .; to Technical Specifications Resulting from 10 CFR 20 and ; 50.36a Changes" (from Christopher I. Grimes to Owners Groups Chairs). This change is consistent with traveler TSTF-152. 5.6-5 DCPP LTOP arming and PORV lift settings are_referged in
~
Yes Yes Yes Yes PTLR section perWWsTF-233) l 7 S.0-803 } 5.7-1 Ar Yes Yes Yes Yes 5.7-2 Chihges"unauthor " to "inadvert t" in High , W W W Ves- pg y_q Radiation Are tion to refl the sposip6nas
- stated 1 8.3.8. Section . r ngphyytfal lbarr s for High Radia Area This change is ,
sistent with tra TSTF- 7. Id5ECT* 69-Sc.y 5.7-3 This change deletes the phrase "or that cannot be No No No Yes continuously guarded" from the ITS for Callaway to make - them consistent with the CTS. INSEUl" Q5.1-I] . I i b* *
- i r
i WCGS-Conversion Comparison Table-ITS5.0 MMI
INSERT 6B-Sa 0 5.2-1 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 5.6-6 The ITS requirement to provide Yes Yes Yes Yes documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves is deleted. INSERT 6B-5b 0 5.2-1 TECH SPEC CHANGE APPLICABILITY , DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER Yes Yes Yes 5.7-1 Section 5.7 is revised in accordance with Yes' 10 CFR 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. INSERT 6B-Sc 0 5.2-1 TECH SPEC CHANGE APPLICABILITY-DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION Yes Yes Yes Yes 5.7-2 ITS 5.7.2.e is revised consistent with CTS l 6.12 that allows any individual or group of individuals to enter a high-high radiation area (dose rates greater than 1.0 rem / hour at 30 cm) accompanied by an individual qualified in radiation protection procedures with a radiation dose rate monitoring device.
e INSERT 6B-5d 0 5.2-1 TECH SPEC CHANGE APPLICABILITY > NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY Yes Yes 5.7-4 ITS 5.7.2.f is revised consistent with CTS Yes Yes 6.12 to delete the phrase "that is controlled as a high radiation area". The , proposed change would preclude having to post an area around the high-high radiation area as a high radiation area when the area may not meet the definition of a high 1 radiation area. t i I L i f a 4 _ . _ _ . _ ._ ___ ._.__._____m ___.___._...__._____..____.._m..____m____._s___ _____ _ _ _ ____m__._ _ _ _ _ _ _ _ . _ _ .___._______.________m_.__ ___m.__ . . .___ -..m__._____ _ _ _ ____
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.3-1 APPLICABILITY: DC, WC, CA REQUEST: ITS 5.3.1 (Wolf Creek, Callaway and Diablo Canyon) Comment: Part 55 of Title 10 of the Code of Federal Regulations was revised in March 1987 to establish upgraded requirements for licensed reactor operators. NRC Regulatory Guide (RG) 1.8, Revision 2, April 1987, describes methods acceptable to the staff for complying with the revised rule. The Statements of Consideration for the Part 55 rule change state that, "Those facility licensees that have made a commitment that is less than that required by the new rules must conform to the new rules automatically " The staff is concerned some facilities continue to have technical specifications that reference older industry standards that may not fully meet the revised requirements of 10 CFR Part 55. l The staff previously considered that the standards applied through the industry's accreditation process were equivalent to the guidance contained in RG 1.8, Revision 2. However, the staff ' has recently found that current INPO guidance in this area is very general; only advising licensees to follow regulatory requirements, in RG 1.8, Revision 2, the NRC staff endorses, with conditions, certain parts of industry standard ANSI /ANS-3.1-1981 as an acceptable approach for complying with the qualification and training requirements of 10 CFR Parts 50 and 55. This endorsement applies to the positions identified as shift supervisor, senior operator, licensed operator, shift technical advisor, and radiation protection manager. For positions other than those identified, the RG finds acceptable the approach provided in ANSI N18.1-1971. For Callaway, the ITS proposes to adopt the CTS which adopts ANSl/ANS 3.1-1978 for the unit staff (besides SROs, ROs and STAS) and RG 1.8, September 1975 for the radiation protection manager. For Wolf Creek, the ITS proposes to adopt the CTS which adopts ANSI /ANS 3.1-1978 for the unit staff (besides SROs and ROs) and RG 1.8, September 1975 for the radiation protection manager. For Diablo Canyon, the ITS proposes to adopt the CTS which adopts ANSI /ANS 3.1-1978 for the unit staff (besides the radiation protection manager) though it does makes a reference to ROs and SROs having to meet the minimum qualifications of Part 55. Please describe how your commitment to an ANSI standard other than that endorsed by NRC RG 1.8, Revision 2 currently meets the requirements of 10 CFR Part 55, as discussed in the Statements of Consideration for the rule change and would meet those requirements with the iTS as proposed. FLOG RESPONSE: Callaway, Wolf Creek, and Diablo Canyon believe that this question is outside the scope of the ITS conversion process because it is a generic industry question. The NRC's question regarding compliance with 10 CFR Part 55 shnuld be discussed on a generic basis. Therefore, it is proposed that this issue be resolved separate from the ITS conversion and the submitted ITS 5.3.1 remain as is which is consistent with the CTS. ATTACHED PAGES: None l l
l l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 5.5-2 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 5.5-14 l l Comment: WOG-85 has not yet become a TSTF. Use current ITS. FLOG RESPONSE: WOG-85 has been approved by the TSTF and is designated as TSTF-237. This traveler has been submitted to the NRC and is under review. > The proposed wording in TSTF-237 was modified from WOG-85 and l these modifications have been incorporated into the ITS. The FLOG
- continues to pursue the changes proposed by this traveler.
For Wolf Creek, this change was approved by the NRC in Amendment ; I No.106 dated June 24,1997. Therefore, the wording in ITS 5.5.7 is consistent with Amendment No.106. , ATTACHED PAGES: Encl.5A Traveler Status page Encl. 6A 3 l Encl. 68 4 l l 1 l l l I i l
I i I l INDUSTRY TRAVELERS APPLICABLE TO SECTION 5.0 l TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated NRC approved. TSTF-37, Rev.1 Incorporated 5.6-2 DCPP onlyi ___ TSTF-52 Incorporated 5.5-4 h@#'fd.Qh4 q 5.u -6 l TSTF-65 @ Incorporated NA MNRC approv ag w s]
@ s.z-9
<atweie,4twra _g TSTF-106, Rev.1 Not Incorporated NA Retain CTS.
TSTF-118 Incorporated 5.5-8 %epr@[Tcr.o-m(.)
@TTi1V jieflacpqforatV A /TetsjeCJ}8 [re.c.o-we. J TSTF-120@ Not Incorporated NA Retain CTS Prn s.o-oer. _\
Morfo_g_ eV- / 52-F7 !o s.2-L) TSTF-152 Incorporated 5.6-4 %'appr@) ATRco -co s-) (TMfAtIP ./Jacocs#rasetf / 7 St7,r) Fo 5 2-'l l
%M7 Incorporated 5.6-5 hkro@4 rR.s.o-cu31 l
WOG-72 Incorporated 5.5-13 I# " * \ Incorporated 5.5-14 bepkIrnaler Incorporated s 4-2, 5.k s.2 .5, upac)i6n pg,g., i ' rsrs=- 1 5 8 ff,f;l:f ,2,5 C-% e 4. i o S/lS/97
l CM NUPEER JUSTIFICATION 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies." to provide consistency with current application of these requirements. This is consistent with the use of current TS and alleviates potential confusion in the program descriptions. This change is consistent with traveler TSTF-118. 5.5 9 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). 5.5 10 Section 5.5.12c specifics a surveillance program to ensure that the quantity of radioactivity contained in all outside liquid radwaste tanks that are not surrounded by liners, dikes, .or walls is less than the predetermined quantities. This change lists the tanks that the surveillance program is applicable to as is in the current TS. This change is a plant specific requirement consistent with the current TS. 5.5 11 The documents referenced for the testing frequency for the Ventilation Filter Testing Program (VTFP) do not provide frequencies for combined pressure drop tests or the heater power rating test. The current TS frequency is added for these two tests. 5.5 12 Not applicable to WCGS. See Conversion Comparison Table (Enclosure l 68). 5.5 13 This change revises Radioactive Effluent Controls Program dose projections to meet original intent of TS prior to implementation of GL 89 01. GL 89 01 provided the wording for the STS (Section 5.5.4.e) ' which combined the requirements for cumulative and projected dose. This requires a plant to make projected doses for the quarter and year on a 31 day basis. It is only necessary and reasonable to make a projection for the next 31 days. A cumulative dose projection is still l required for the current calender quarter and year in accordance with-L the ODCH. This change is consistent with traveler WOG 72. l T5TF-23~7 6 5.5 14 ITS Section 5.5.7__is being revised consistent with traveler h [and Ci m ite amendiFent Sequest -maitted ? Mm). The proposed $mendmen+ No. changes to ITS 5.5.7 provide an exception to the examination
& dated June."14,FM7 requirements in Regulatory Guide 1.14. Revision 1, " Reactor Coolant Pump Flywheel Integrity." The proposed exception to the l
l recommendations of Regulatory Position C.4.b would allow for an I acceptable inspection method of either an ultrasonic voltmetric or surface examination. The acceptable inspection method would be conducted at approximately 10 year intervals. This change is consistent with the NRC Safety Evaluation Report associated with [ WCAP 14535, " Topical Report on Reactor Coolant Pump Flywheel Inspection
- Elimination."
WCGS-DsfferencesfromNUREG-1431-ITS5.0 3 S/l5/97
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 4 of 5 SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUtEER DESCRIPTION No No No 5.5-12 The referenced frequencis for the tests listed in the Yes Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96-09) Yes Yes Yes 5.5-13 Revises Radioactive Effluent Controls Program dose Yes projections to meet original intent of TS prior to implementation of GL 89-01. (WOG-72) Yes Yes Yes 5.5-14 Section 5.5.7 is being revsied consistent with traveler Yes [and(al$cEng .$.-MtD@esvfubstettd pacenterv %g 4g37 goc, -p 3,g,g The proposed changes to Section 5.5.7 provide g3 q g.y en exception to the examination requirements in Regulatory - Guide 1.14. Revision 1.
- Reactor Coolant Pump Flywheel Integrity."
No No Yes 5.5 15 This change provides a time interval of within 31 days No after removal in which a laboratory test or a sample obtained from the charcoal adsorber must be tested. This change is consistent with Callaway CTS. Yes. LAR 94-14 No. Wolf Creek CTS No. Callaway CTS 5.6-1 Revises Section 5.6.4, "Honthly Operating Report." to No. DCPP CTS consistent with consistent with consistent with reflect a revised submittal date. NUREG-1431. NUREG-1431. NUREG-1431. Yes No. Not in CTS No. Not in CTS. No. Not in CTS. 5.6-2 Deletes the EDG Report to reflect the recomumendations of equir e f r D --.jd? s.4.-2. ] _dat ay 31 r5TF- 37, Rev i . _ Yes. Consistent Yes. See LA 42/28. Yes Yes 5.6-3 Revises report dates in ITS 5.6.2. " Annual Radiological Environmental Operating Report" to be consistent with with CTS and LA current TS. 78/77. (5. N- IC. IMSER.T G8 -4 a}l o s.2.-t ) lu.%#.T 64 - 4bM c:A S.o-otG l (5.s-18 t use R.T r.t> - 4 c_}} pc s.o - cos [ S/1S/97 WCGS-Conversion Comparison Table-ITS 5.0
l l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-3 APPLICABILITY: DC, CP, WC, CA , l REQUEST: ITS 5.5.4 b&g and Difference 5.5-1 l 1 l Comment: Changes are based on a yet unnumbered traveler. Use current ITS. I FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. This l traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the April 9,1997 letter I from C. Grimes (NRC) to J. Davis (NEI), with some exceptions. The FLOG submittals have been revised to incorporate TSTF-258. The latest industry status on TSTF-258 is that the NRC has requested changes to Section 5.7, High Radiation Area. See response to Comment Number l 5.7-1 for how the FLOG has addressed the NRC comments on TSTF-258. l l ATTACHED PAGES: See markups associated with Comment Number Q 5.2-1. l
I l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 5.5-4 APPLICABILITY: DC, CP, WC, CA 1 REQUEST: ITS 5.5.4 e and Difference 5.5-13 i i Comment: WOG-72 has not yet become a TSTF. Use current ITS. FLOG RESPONSE: This change to ITS 5.5.4 e was prepared in accordance with WOG-72, Rev.1 which is currently under TSTF review. The change specifies the' the requirement to determine cumulative dose contributions from radioactive effluents need be done on a current quarterly and annual basis instead of every 31 days. We believe there is a strong technical basis for this change to the iTS. We request that the NRC keep this as an open item under the assumption that the traveler will be approved prior to issuance of the SER. ATTACHED PAGES: None
1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 5.5-7 APPLICABILITY: WC REQUEST: CTS 3.7.6 and Changes 10-15-LG and 10-17-A (Wolf Creek) l Comment: The CTS markup is inconsistent with the comments as nothing is lined out. . . Further, the deletions (at least as they are reflected in ITS 5.5.11) need a better explanation. Provide explanation. I FLOG RESPONSE: The CTS markup should have lined out references to the overall Pressurization System flow rate of 2,200 cfm i 10% in two places, i.e., in CTS 4.7.6.c 1) and CTS 4.7.6.c 3). As further justification for DOC 10-15-LG, the following paragraph is added: "The acceptability of deleting the overall Pressurization System flow rate is based on the system design. The design flow rate for each Pressurization System fan is 2,200 cfm as shown on FSAR/USAR Table 9.4-4. Air flow into the fan consists of the flow from the Pressurization System filter adsorber unit plus a much larger flow of recirculated air from the Control Building. Thus, the 2,200 cfm flow rate through the system has only an indirect effect on the flow rate through the filter adsorber unit. The relative flow rates from the filter adsorber unit and from the Control Building are established by performing a flow balance on the Pressurization System. The proposed surveillance testing in accordance ; with the Ventilation Filter Testing Program (ITS 5.5.11) will assure the required flow rate through the Pressurization System filter adsorber unit. 3 Based on the above discussion, there is no need to specify an overall system flow rate in the ITS." While evaluating this question, the need to further modify ITS 5.5.11 for WCGS was identified. The basis for this change is that CTS 4.7.6c.3) and . 4.9.13b.3) requires measuring flow rate while imposing an artificial differential pressure (dP). This requirement is unique to WCGS among the FLOG plants. Therefore, this CTS requirement has been incorporated into the ITS as Section 5.5.11.f. New JFD 5.5+19 is initiated to address adding this requirement to the CTS. ATTACHED PAGES: Attachment No 13, CTS 3/4.7 - ITS 3.7 Encl. 2 7-15 Encl. 3A 13
- Attachment No.17, CTS 6.0 - ITS 5.0 Encl. 5A 5.0-23 Encl.6A 4 Encl. 6B 4
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (New) Perform required CREVS filter testing in accordance with the Ventilation #'510 08-AN Filter Testing Program (VFTP);
- c. At least once per 18 months or (1) after any structural maintenance ""' 10-08-A^9
' - ~ - ~ "
on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the Control Room Emergency Ventilation System satisfies the in-place penetration and bypass leakage testing
$1100$f@d acceptance enteria; of less than 1% for HEPA filters and 0.05%
for charcoal adsorbers and uses the test procedure guidance in Regulatory Positions C.5.a. C.S.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is jo s,5 -7] 2000 cfm t10% for the Filtration System and(2290xAnGae3 Wet _ i1N5-1Xi% f"MENA4 f tt)>P ostumr.atsorusystemes)750 Pressunzabon System filter adsorber unit; cfm 110% gpmgpfrog@the
- 2) Verifying, ^5 21 d:;: :": ::=0ve!, that a laboratory ' 10-23-LS-13 ' '
' '~~ - ~ ~+
analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, 10 08-A^ 9
' "'~~' -^^ "
Revision 2, March 1978 meets the laboratory te ' cnteria of ASTM D3803-1989 when tested at 30* C and relative q 3,g.42. I humidity, for a methyl iodide penetration of less than 2%; and
- 3) Verifying system flow rate of 2000 cfm 110% at greater than "1017-A " 7
- 4 Mr eaual to 6 6 inches W G fdirty filter) for the Filtration hystem andG20Q.sMP499@t greater than or equal to 3.6_ % s.5-7 } ' 10-15-LG'l
' " ^ ^ ~ ' ' " ' ' '
inches W.G. (dirty filter) for the[PJessurtzJett005ValemM _ 7JGtfin4TOMrtom71broQgh the)Pressunzation System filter i adsorber unrt dunng sysiem operation when tested in accordance with ANSI N510-1980.
- d. After every 720 hours of charcoal adsorber operation by verifying 110-23-L5%!31
""'M 21 d:';: :".0 :m0v:! that a laboratory analysis of a represen- " " *
- tative carbon sample obtained in accordance with Regulatory Position M"
il008WP ~ N" C,6.b of Regulatory Guide 1.52 Revision 2. March 1978, meets the laborato testing enteria of ASTM D3803-1989 when tested at 30'C and relative humidity, for a methyl iodide penetration ofI s than 2%; p~7M, -} t$ 5.S-2 }
- e. At least once per 18 montns by:
- 1) Verifying that the pressure drop across the combined HEPA ' {l0 0$lAN ]
filters and charcoal adsorber banka is less than 6.6 inches Water Gauge while operating the system at a flow rate of 2000 cfm 110% for the Filtration System and less than 3.6 inches Water Gauge while operating the system at a flow rate of 750 cfm 110% for the Pressurization System filter adsorber unit
- 2) Verifying that on(an actual or simulated actuatioit: Centre! o.com (1010-TR;1{ .
Venti!:t!On ! r'Meh c: We Ortreur Redir:Mt; tett signal, the
- system automaticallyfactuateskef!!:her Mte recircu!:t!cn mede Of 0;' erat!On
""" *~ff 'hreu;h '50 HEP ^ C!!er: and chcrrer' ed:Oter b:M WOLF CREEK - UNIT 1 3/4715 Amendment No. 32-102 Mark-up of CTS.iN,7 SAS/97
_ _ ~ _ _ l 1
- CHANGE HUfE8 N21C DESCRIPTION 10 09 LS 27 This change deleted the ACTION for an OPERABLE ventilation train not being capable of being supplied from an emergency power source. Per the definition of OPERABLE in NUREG 1431 the ventilation system would be considered OPERABLE with either a NORMAL or EMERGENCY power source. l i
10 10 TR 1 The SR is revised to allow credit for an actual actuation, if i one occurs, to satisfy the SRs. The identification of the ' initiating signal is moved to the Bases. , i 10 11 LS 19 The frequency of the surveillance requiring verification of the CR ventilation system capability to maintain a positive pressure is relaxed to 18 months on a STB consistent with NUREG 1431. The new frequency requires one of the two trains to be tested every 18 months instead of both trains every 18 months. The most likely cause of a failure to achieve the required pressure is a failure of the ventilation pressure boundary. Thus when one train successfully demonstrates the ability to maintain the pressure, in all likelihood the other train will also. This results in less testing of the CR ventilation system than is required by the CTS. 10 12 LS 32 This change deletes the required STB for the 31 day test. Both trains will still be tested on a 31 day frequency. This change is acceptable based on the evaluation of the - effectiveness of STB testing provided in NSHC LS 32. 10 13 LG Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 10 14 A The statement that LCO 3.0.4 is not applicable is deleted based upon the new ITS definition of LC0 3.0.4 which does not apply in MODES 5 and 6. 10 15 LG The ventilation system flow rates would be moved to licensee controlled documents. These flow rates are established in con.iunction with flow balancing of the ventilation systems. (_INSe y _3A-13 C -lcp es. 5-7 l 10 16 LG Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 10 17 A The SR to measure ventilation system flow rate is not identified as a separate SR in the ITS because it is verified during the other in place filter tests (see ITS 5.5.11) 10 18 LS-36 Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 10-19 A Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). WCGS-Description of Changes to CTS 3H. 7 13 S/15/97 I
l INSERT 3A-13a 0 5.5-7 10-15-LG The acceptability of deleting the overall Pressurization System l flow rate is based on the system design. The design flow rate for each Pressurization System fan is 2,200 cfm as shown on FSAR/USAR Table 9.4-4. Air flow into the fan consists of the flow from the Pressurization System filter adsorber unit plus a l much larger flow of recirculated air from the Control Building. ' Thus, the 2,200 cfm flow rate through the system has only an indirect effect on the flow rate through the filter adsorber unit. The relative flow rates from the filter adsorber unit and from the Control Building are established by performing a flow balance on the Pressurization System. The proposed surveillance testing in accordance with the Ventilation Filter l Testing Program (ITS 5.5.11) will assure the required flow rate I through the Pressurization System filter adsorber unit. Based on the above discussion, there is no need to specify an overall system flow rate in the ITS.
Programs and Manuals 5.5 5.5 Programs and Manuals
- d. Demonstrate at'ieast once per 18 monttis for each of the ESF IEE5#1121 systems that the pressure drop across the combined HEPA filters.
~' ' ' ~ '~
j the prefilters, and the charcoal absorbers is less than the I value specified below when tested in accordance with Regulatogy ) GQideE 52ERuvision 2, er.d AS",C 5 10 1000] at the sysiem - $NN. l flowrate specified below ft 10t3 ESF Ventilation System Delta P Flowrate beq+*cy %%%m Qte _y I u.c s .o -oo < ] Control Room. iltration:@ 6.671n. W.' G: 2000;cfm NR6?PS%d ! 3:6 fin..y. G~. 75DJ,cfm Control Roosh, Press.urization Aux 111ary/Fthl'BuiTding Emergency @ Exhaust 4:7~ in. - W.' G'.~. u 165001cfm M&Pb b M od bs c. S .O - oo b]
- e. Demonstrate at least once per 18 months that the. heaters for $1515M13 each of Lthe ESF' systems dissipate the value specified below:when tested in accordance with ASME N510-1975.~ ~ 1. ' ~ lR 28iPS$ x ESF Ventilation System _ Wattage e.mesq.$_% % 6 v on buunuhrb Control Room 5 i;1~.kW gggypSp Auxiliary / Fuel Building 37 i 3 kW f
(mapyEp%} \ The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
+. Nsant s.o-2@ o s.s-7 )
5.5.12 Exolosive Gas and Storace Tank Radioactivity Honitorino Procram This program provides controls for potentially explosive gas mixtures IIBiPST contained in the ' Waste Gas Holdup System, the quantity of , ggy radioactivity contained in gas storage tanks or fcd into th; offi;as ~ trcat,T. cat systcm, and the quantity of radioactivity contained in !$BiPST unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch gg;p3;
~"
Technical Position (BTP) ETSB 11-5. Revision 0, July 1981,
" Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in l accordance with Standard . Review Plan, Revision 2, July 1981, l Section 15.7.3, " Postulated Radioactive Release due to Tank Failures."
1 (continued) i, WCGS-Mark-up ofNUREG-1431-ITS 5.0 5.0 23 5/1587 l
INSERT.S.0-23 0 5.5-7 . i
- f. _ Demonstrate at least once per 18 months for each of the ESF systems that '
following the creation of an artificial Delta P across the combined HEPA filters, the prefilters. and the charcoal absorbers of not less trian the value specified below (dirty filter conditions), that the flowrate j through these flow paths is with i 10% of the value specified below when ; tested in accordance with ANSI N510-1980. , ESF Ventilation System Delta P Flowrate , i Control Room Filtration System 6.6 in. W.G. 2000 cfm - Control Room Pressurization System 3.6 in. W.G. 750 cfm Auxiliary / Fuel Building Emergency Exhaust 4.7 in. W.G.
~
6500 cfm - 1 l
s .s -e s 6O ff a 6 m. A bws 6 p s T4M&de }ec,s.o.wi} HANGE ' %i usinr dA-4Mcp s.5 ~l } l NUMBER JUSTIFICATION 5.5 15 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). 6.s -u, _ t Tass a.r e.* --h) o s. z -i 1 l 5.6 1 Not applicable to WCGS. See Conversion Comparison Table (Enclosure
~
s s;n N A.~r (. A- 4-h -M 5.6 2 This change deletes the Emergency Diesel Generator Report to reflect the recommendations of GL 94-01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. 5.6-3 This change revises the report date in Section 5.6.2, " Annual Radiological Environmental Operating Report," to be consistent with current TS . 5.6 4 This change revises Sections 5.6.1 and 5.6.3, " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report " respectively, per NRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From , Christopher I. Grimes to Owners Groups Chairs). This change is consistent with traveler TSTF 152. 5.6 5 [ ] PORV lift settings are referenced in the PTLR section ( M -1D F-2 7 -m3l Cs. c. - ro lussair G A -4 bH os.t.: _I - 5.7 1 (This 'ang evise High adia nA J ta to pi(orpor chpasf 7 st with 0 CF 0. 1] ./Specifffally, 1 stances fro ( the co tadia on so' ce are at . t use.a.T 6A -4 c. j
~
5.7-2 (Th yi 'c er ses " naut riz 't "in vert t" nteH
' J4di on a se on ref ct e NR s iti n a sta ed 'n /
R ato Gui 8.38 Sec ' n 1 reg din phy cal arr er forAligh dia on Ar s. T si cons tent ith av er T F- 67. l usEcr 6A- 4 ck-5.7 3 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). (54e}[ o G.2-t1 WCGS-Differencesfrom NUREG-1431 -ITS 5.0 4 S/158 7
l INSERT 6A-4a 0 5.2-1 5.5-16 The Radioactive Effluents Controls Program is revised to include clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. These statements of applicability clarify the allowance for surveillance frequency extensions and allowance to perform missed surveillances. Generic Letter 89-01, " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications and the Relocation of Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents Control Program in the Administrative Controls Section of the Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 89- j
- 01. Since this change adopts previous CTS requirements, it is l considered a change of presentation method only. This change is consistent with TSTF-258.
INSERT 6A-4b 0 5.2-1 5.6-6 The ITS requirement to provide documentation of all challenges I to the pressurizer power operated relief valves or pressurizer i safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694, "THI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states:
" Assure that any failure of a PORV or safety valve to close will l be reported to the NRC promptly. All challenges to the PORVs or l safety valves should be documented in the annual report." NRC l Generic Letter 97-02, " Revised Contents of the Monthly Operating Report" requests the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program and availability and capacity statistics. The generic letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent with TSTF-258.
! INSERT 6A-4c 0 5.2-1 l Section 5.7.is revised in accordance with 10 CFR 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. These changes are consistent with the draft Generic Letter (93-XX) on proposed changes to STS NUREGs based on the new 10 CFR 20 and the letter from C. Grimes, NRC. to l J. Davis, NEI dated - April 9,1997. This change is consistent with TSTF-258 and encompasses the NRC comments on 6/11/98. Additional technical changes made to Section 5.7 are identified and justified.
INSERT 6A-4d 0 5.2-1 ITS 5.7.2.e is revised consistent with CTS 6.12 that allows any individual or group of individuals to enter a high-high radiation area (dose rates greater than 1.0 rem / hour at 30 cm) accompanied by an individual qualified in radiation protection procedures with a radiation dose rate monitoring device. The qualified individual is responsible for providing positive control and shall perform periodic radiation surveillances at the frequency specified in the RWP. The CTS requirements allow the qualified individual to enter a locked high radiation area with plant workers without first having to enter the area to determine dose rates and then exit the area to provide dose rate information to the plant workers and then reenter the area. This flexibility is in keeping with the "As Low As Reasonably Achievable" principle while maintaining appropriate radiation worker practices. INSERT 6A-4e 0 5.2-1 ITS 5.7.2.f is revised consistent with CTS 6.12 to delete the phrase "that is controlled as a high radiation area". The proposed change would preclude having to post an area around the high-high radiation area as a high radiation area when the area may not meet the definition of a high radiation area. INSERT 6A-4f CA 5.0-003 ITS Section 5.5.10 is being revised consistent with CTS 8.8.4.c. The proposed change deletes the phrase "and low pressure turbine disc stress corrosion
-cracking" from the ITS to be consistent with the practices of the CTS which do not have this requirement for the Secondary Water Chemistry Program.
INSERT 6A-4g 0 5.5-7 5.5-19 CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety-related ventilation system filter adsorber units include the requirement to measure flow rates within specified values, while imposing an artificial differential pressure, during system operation, when tested in accordance with ANSI N510-1980. This flow rate testing is to be performed at least once per 18 months, after any structural maintenance on the HEPA filter or charcoal adsorber housings, or following painting, fire. or chemical release in any ventilation zone communicating with the system.
. Therefore, the CTS surveillance requirements are incorporated into the ITS.
i l l l
i i
~
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 4 of 5 SECTION 5.0. f i DIFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUleER DESCRIPTION I No No 5.5-12 The referenced frequencis for the tests listed in the Yes No Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96-09) Yes Yes Yes 5.5-13 Revises Radioactive Effluent Controls Program dose Yes I projections to meet original intent of TS prior to implementation of GL 89-01. (WOG-72) t Yes Yes < 5.5-14 Section 5.5.7 is being revsied'ccesistent with traveler Yes __ Ye_s. and(aAh IMtJie$eN Member
- M M Ienf ME to?. }QS.5 ~2.\ I j
The proposed changes to Section 5.5.7 provide gg g : anexceptiontothaexaminationrequirementsinRegulatoryj i Guide 1.14. Revision 1, " Reactor Coolant Pump Flywheel Integrity." No No Yes ! 5.5-15 This change provides a time interval of within 31 days No after removal in which a laboratory test of a sample i obtained from the charcoal adsorber must be tested. This change is consistent with Callaway CTS. Yes. LAR 94-14 No. Wolf Creek CTS No. Callaway CTS l 5.6-1 Revises Section 5.6.4, " Monthly Operating Report." to No. DCPP CTS consistent with consistent with consistent with ; reflect a revised submittal date. NUREG-1431. NUREG-1431. NUREG-1431. No. Not in CTS Nc. Not in CTS. No. Not in CTS. ) 5.6-2 Deletes the EDG Report to reflect the recommendati s of Yes
-ias.d.-2.l
~
f D J iat 31 TsTF- 3 7, Re.v l . Yes i Revises report dates in ITS 5.6.2. " Annual Radiological Yes. Consistent Yes. See LA 42/28. Yes 5.6 3 I Environmental Operating Report" to be consistent with with CTS and LA current TS. 78/77. s.5-lat~ iwAsiav ~ s ro .f-@)es.s i { ' (S.5 - 4 NsER.T sa-4 a.}l o s.a.i ) kE lobiit.T 64 - 4%ah\ CA S.O-OtXh G s-ilHTGeram i="T *-+o-t x s o =2 I j SM5M7 nrinn remnariwn Tahlr 115 % n l
t 4 INSERT 6B-4a 0 5.2-1 i TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION-Yes Yes Yes Yes. 5.5-16 The Radioactive Effluents Controls Program is revised to include clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. t INSERT 6B-4b CA 5.0-003 , TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION Yes Yes 5.5-17 This change deletes the phrase "and low pressure Yes No , turbine disc stress corrosica cracking" from ITS ; 5.5.10 to make the program consistent with CTS , 6.8.4.c. DC 5.0-003 INSERT 6B-4c TECH SPEC CHANGE APPLICABILITY a DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION Yes No No No 5.5-18 Revises DCPP Sections 5.5.9 and 5.6.10 to ' reflect License Amendment 124/122. dated March 12, 1998, which allows implementation of steam l generator tube voltage based on repair criteria for ODSCC indications at tube to tube support . ! plate intersections. , L i i y P t 1
INSERT 6B-4d 0 5.5-7 TECH SPEC CHANGE APPLICABILITY DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER No No Yes No 5.5-19 Wolf Creek CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety-related ventilation system filter adsorber units include the requirement to measure flow rates within specified values, while imposing an artificial dif ferential pressure, during system operation, when tested in accordance with ANSI N510-1980. The CTS surveillance requirements are incorporated into the ITS.
ADDITIONAL INFORM / 'lON COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-8 APPLICABILITY: DC, CP, WC, CA REQUEST: CTS 3.7.6 (3.7.5.1 and 3.7.6.1 - DCPP and 3.7.7.1 and 3.7.8 -CPSES) and Change 10-08-A Comment: It should be specifically noted as to which CTS requirements were carried over to the VFTP and which were deleted (as well as which section of what standard justified the duplication deletions). Provide explanation and justification. FLOG RESPONSE: Attached Table 5.5-8 describes where the CTS SRs for plant ventilation systems were moved to in the ITS. The following provides justification and clarification for those CTS SRs that were not moved to either the
" Ventilation Filter Testing Program (VFTP)" in the ITS or the ITS SRs:
. DOC 10-07-LG (not applicable to CPSES) moves the requirement to verify Control Room temperature once every 12 hours to a licensee controlled document. This DOC has been revised to include the following additional justification: "The NRC has previously approved moving this type of detailed information or specific requirements to a licensee controlled document that is maintained in accordance with applicable regulatory requirements. This temperature is not an initial condition or controlled parameter for any licensing-based accident scenarios Also, its inclusion in the ITS is not necessary to adequately protect the health and safety of the public. The basic requirements for maintaining OPERABILITY are still retained in the technical specifications."
. Per DOC 10-17-A, the SR to measure ventilation system flow rate is not identified as a separate SR in the ITS because it is verified as part of the other in-place filter tests that are specified in ITS 5.5.11. The same DOC applies to CTS SR 4.7.6.1 b 3 for Diablo Canyon, CTS SR 4.9.13 b 3 for Wolf Creek and CTS SR 4.7.7 b 3 for Callaway for the same reason.
. DOC 10-08 A has been revised to show that some CTS SRs were moved to the ITS SRs.
ATTACHED PAGES: Attachment No.13, CTS 3/4.7 - ITS 3.7 Encl. 3A 12 Encl. 3B 13
. -- .- _ . . . - - - - = - .- - . . .
TABLE QS.5-8 DCPP WC CA CP Licensee
. CTS SR . . CTS SR CTS SR CTS SR Controlled VFTP iTS SR Document 4.7.5.1 a 4.7.6 a 4.7.6 a N/A X 4.7.5.1 b 1 4.7.6 b 4.7.6 b 4.7.7.1 a 3.7.10.1 4.7.5.1 b 2 N/A N/A N/A 3.7.10 Bases 4.7.5.1 b 3 N/A N/A N/A 3.7.10 Bases 4.7.7.1 b ITS 5.5.11 3.7.10.2 4.7.5.1 c 1 4.7.6 c 1 4.7.6 c 1 4.7.7.1 b 1 ITS 5.5.11a&b 4.7.5.1 c 2 4.7.6 c 2 4.7.6 c 2 4.7.7.1 b 2 ITS 5.5.11c 4.7.5.1 c 3 4.7.6 c 3 4.7.6 c 3 4.7.7.1 b 3 See DOC 10-17-A ,
4.7.5.1 d 4.7.6 d 4.7.6 d 4.7.7.1 c ITS 5.5.11 & 3.7.10.2 l 5.5.11c l 4.7.5.1 e 1 4.7.6 e 1 4.7.6 e 1 4.7.7.1 d 1 ITS 5.5.11d 3.7.10.2 4.7.5.1 e 2 4.7.6 e 2 4.7.6 e 2 4.7.7.1 i 3.7.10.3 4.7.5.1 e 3 4.7.6 e 3 4.7.6 e 3 4.7.7.1J 3.7.10.4 4.7.5.1 e 4 4.7.6 e 4 4.7.6 e 4 4.7.7.1 d 2 ITS 5.5.11e 3.7.10.2 4.7.5.1 f 4.7.6 f 4.7.6 f 4.7.7.1 e ITS 5.5.11 & 3.7.10.2 5.5.11a l 4.7.5.1g 4.7.6 g 4.7.6 g 4.7.7.1 f ITS 5.5.11 & 3.7.10.2 l 5.5.11 b 4.7.7.1 g ITS 5.5.11 & 3.7.10.2 5.5.11a 4.7.7.1 h ITS 5.5.11 & 3.7.10.2 5.5.11b 4.7.6.1 a 1 4.9.13 a 4.7.7 a 4.7.8a 3.7.12.1 DC&CP 3.712.1 Bases 3.7.13.1 WC&CA 4.7.6.1 a 2 N/A N/A N/A 3.7.12.1 Bases 4.7.8b ITS 5.5.11 3.7.12.2 4.7.6.1 b 1 4.9.13 b 1 4.7.7 b 1 4.7.8 b 1 ITS 5.5.11a&b 3.7.12.2 DC 3.7.13.2 WC&CA NA
-CP 4.7.6.1 b 2 4.9.13 b 2 4.7.7 b 2 4.7.8 b 2 ITS 5.5.11c 3.7.12.2 DC 3.7.13.2 WC&CA NA-CP 4.7.6.1 b 3 4.9.13 b 3 4.7.7 b 3 N/A See DOC 10 A 4.7.6.1 c 4.9.13 c 4.7.7 c 4.7.8 c ITS 5.5.11 & 3.7.12.2 DC&CP 5.5.11c 3.7.13.2 WC&CA 4.7.6.1 d 1 4.9.13 d 1 4.7.7 d 1 4.7.8 d 1 ITS 5.5.11d 3.7.12.2 DC&CP 3.7.13.2 WC&CA 4.7.6.1 d 2 4.7.7 b 2 4.7.7 d 3 4.7.8 d 2 3.7.12.3 DC&CP 3.7.13.3 WC&CA 4.7.6.1 d 3 4.9.13 d 2 4.7.7 d 4 4.7.8 d 3 ITS 5.5.11e 3.7.12.2 DC&CP 3.7.13.2 WC&CA 4.7.6.1 d 4 N/A N/A N/A 3.7.12.6 3.7.12.6 Bases 4.7.6.1 e 4.9.13 e 4.7.7 e 4.7.8 e ITS 5.5.11 & 3.7.12.2 DC&CP 5.5.11a 3.7.13.2 WC&CA 4.7.6.1 f 4.9.13 f 4.7.7 f 4.7.8 f ITS 5.5.11 & 3.7.12.2 DC&CP 5.5.11 b 3.7.13.2 WC&CA N/A 4.7.7 b 1 4.7.7 d 2 4.7.8 d 4 3.7.13.4 WC&CA 3.7.12.4 CP
CHANGE NUMER RSE DESCRIPTION 09 07 A- A note is added to the [ESW] surveillance that clarifies system operability requirements. Isolation of [ESW) flow to individual components does not render the system inoperable. This change is in accordance with NUREG 1431. Rev.1. and provides clarification only. 10 01 LG Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 10 02 H The APPLICABILITY and applicable ACTIONS are revised to incorporate "during movement of irradiated fuel assemblies" in addition to all MODES (i.e., MODES 1 6). 10 03 LS-7 Not Applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 10 04 A A new ACTION Statement is added by NUREG 1431 to require entering TS 3.0.3 immediately if two trains of the CR ventilation system are inoperable in MODES 1, 2, 3, or 4. The CTS requires entry into TS 3.0.3, since the condition of two trains inoperable is undefined, therefore, the revision has been classified as administrative. 10 05 LS 18 A new option is added to the ACTION by NUREG 1431 that allows the suspension of CORE ALTERATIONS or movement of irradiated fuel versus placing the CR ventilation system in the recirculation mode. 10 06 LG The details and description of the required actions and the monthly SRs for train operability are relocated to the Bases. This is an example of removing details that are not required to be in TS and is consistent with NUREG 1431 Rev. 1. 10 07 LG 3'1 t doc . NSEst T' 34 -12 c. - - - Q s.5-8 [ 10-08 A The description of the ventilation filter specific testing requirements and the required surveillances are moved to the Ventilation Filter Testing Program (VFTP) as defined in the Administrative Controls of the ITS. No technical changes to requirements or test specifics except as noted in separate change numbers are made. A new SR is added that requires [CREVS and Emergency Exhaust System] ventilation system , g filter testing in accordance with the VFTP. The requirements of this specification ar#: 1) moved to Section 5.5.11_ot the ITS. or 2) d tney a ca 1 at6
.R 1sioV2. [ SI 10 1
__ 0. D 803 9 8'8'* l kmoved io ITS SR3) WCGS-Description of Changes to CTS 3/4.7 12 $/15/97
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. _ . _ _ _ . _ _ _ _ _ _ . _ -~ _. _ . . _ _ . _ . . ~ . _ . ..- _ _ . _ . _ _ ._._.._. _
l INSERT 3A-12c 0 5.5-8 The surveillance that verifies CR temperature once per 12 hours is moved to a , licensee controlled document. The NRC has previously approved moving this type of detailed information or specific requirements to a licensee controlled document that is maintained in accordance with applicable regulatory requirements. This temperature is not an. initial condition or controlled parameter for any licensing-based accident scenarios. Also, its inclusion in
~the ITS is not necessary to adequately protect the health and safety of the public. The basic requirements for maintaining OPERABILITY are still retained l in-the technical specifications.
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-9 APPLICABILITY: DC, WC, CA REQUEST: CTS 3.9.13 (3.9.12 - DCPP) and Change 12-04-A (Wolf Creek, Callaway and Diablo Canyon) Comment: It appears that some of the CTS requirements covered by this change were deleted rather than transferred to ITS 5.5.11 as stated. Justify the individual deletions. FLOG RESPONSE: Attached Table 5.5-9 describes where the OTS SRs for fuel building ventilation systems were moved to in the ITS. The following provides justification and clarification for those CTS SRs that were not moved to l either the " Ventilation Filter Testing Program (VFTP)" in the ITS or the ITS ' i SRs. Per DOC 12-11-A, the SR to measure FHBVS flow rate is not identified as . a separate SR in the ITS because it is verified during the other in-place I filter tests specified in ITS 5.5.11, " Ventilation Filter Testing Program (VFTP)", and specific ITS SRs. This change does not result in a change to the technical requirements. DOC 12-04-A has been revised to more clearly describe where the CTS SRs were moved to in the ITS. ATTACHED PAGES: i Attachment No.15, CTS 3/4.9 - ITS 3.9 Encl. 3A 8, 9 l l l l l i r
TABLE Q5.5-9 DCPP WC CA CP CTS SR CTS SR CTS SR CTS VFTP ITS SR SR 4.9.12 a 4.9.13 a 4.9.13 a N/A 3.7.13.1 4.9.12 b 1 N/A N/A N/A 3.7.13.5 4.9.12 b 2 4.9.13 b 1 4.9.13 b 1 N/A ITS 5.5.11a &b 3.7.13.2 4.9.12 b 3 4.9.13 b 2 4.9.13 b 2 N/A ITS 5.5.11c 3.7.13.2 4.9.12 b 4 4.9.13 b 3 4.9.13 b 3 N/A See DOC 12-11-A 4.9.12 c 4.9.13 c 4.9.13 c N/A ITS 5.5.11c 3.7.13.2 4.9.12 d 1 4.9.13 d 1 4.9.13 d 1 N/A ITS 5.5.11d 3.7 13.2 4.9.12 d 2 4.9.13 g 1 4.9.13 d 2 N/A 3.7.13.3 4.9.12 d 3 4.9.13 g 2 4.9.13 d 3 N/A 3.7.13.4 DC 3.7.13.5 WC&CA 4.9.12 e 4.9.13 e 4.9.13 e N/A ITS 5.5.11a 3.7.13.2 4.9.12 f 4.9.13 f 4.9.13 f N/A ITS 5.5.11b 3.7.13.2 N/A 14.9.13 d 2 4.9.13 d 4 N/A ITS 5.5.11e 3.7.13.2 WC&CA
l l CHANGE NUMBER ESIE DESCRIPTION irradiated fuel immediately which would establish l conditions outside the Applicability of the LCO. 11 03 M Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 3B). 11 04 LG This change moves the restriction on crane _ operation to a l iicensee_ controlled document 3 ffhe rest tion cra opera sm be r ed cause i is not th as ptio used r th HA. C ne oper ons at co d l , dver y aff fue tored the s f pool l co oll . s ana ed in e review fh y load, l l veme . Th change ' consis nt w h NURE 431, l Rev , and ves regg rements at not the ' q iteria r incluWon in the TS. N SEA.T 3A- Ba- Q S.9 -Z'L l I 12 01 LS 24 The apolicabilaty would be changed to "During movement of irrWRted fuel in the fuel building" instead of "Whenever irradiated fuel is in the spent fuel pool." consistent with NUREG 1431 Rev. 1. The proposed applicability is consistent sith the assumptions used in the Fuel Handling Accident in the fuel building which postulates the inadvertent drop of an irradiated fuel assembly. Potential damage to fuel assemblies due to dropping of heavy loads is addressed by change 12 02 LG. 12 %. U3 Moves the restriction on crane operations over the spent fuel storage areas when the fuel building air cleanup system was inoperable. The restriction on crane operations may be moved because it is not consistent with the assumptions used for the FHA. Crane operations that could adversely affect fuel stored in the spent fuel pool are prohibited in accordance with plant procedures as analyzed in the review of heavy load movements. 12-03 A The statement that 3.0.3 [and 3.0.4] are not applicable would be removed. This is consistent with the proposed change to integrate the emergency exhaust system requirements for irradiated fuel handling in the fuel building with the emergency exhaust systen requirements in Modes 1 tarough 4. ITS 3.7.13 supports this integration of requirements. _ _ Q 4G-1
^ '
t e m" l l Q 3A-e6 WCGS-Description of Changes to CTS 3/4.9 8 5/15/97
INSERT 3A-8a 0 3.9-22 The requirement to suspend crane operations over the spent fuel pool-in the event pool water level is <23 feet, has been removed from the ACTION of CTS 3/4.9.11 (CTS 3/4.9.10 for Comanche Peak) in corresponding ITS 3.7.5, for the fuel pool water level . The bounding design basis fuel handling accident in the spent fuel pool assumes an irradiated fuel assembly is dropped onto [an array of irradiated fuel assemblies seated in) the spent fuel pool. Crane operations that could adversely affect fuel stored in the spent fuel pool are controlled in accordance with plant procedures as analyzed in the review of heavy loads movements. Administrative controls are employed to prevent the handling of loris that have a greater potential energy than those which have been analyzed. Also see' licensees responses to NRC Bulletin 96-02, " Movement l of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment." Moving this information maintains consistency with NUREG-1431. The information is moved to a licensee
-controlled document which is controlled by a 10 CFR 50.59 change j -process.
l INSERT 3A-8b 0 5.5 The description of the ventilation filter specific testing requirements , and the required surveillances are moved to the " Ventilation Filter ! Testing Program (VFTP)" as defined in ITS 5.5.11. No technical changes to requirements or test specifics except as noted in separate-DOCS are made. A new SR is added that requires [ Fuel Building Emergency Exhaust System] filter testing in accordance with the VFTP. The requirements of-this specification are : 1) moved to ITS 5.5.11, or 2) "toved to ITS SRs. l l 1
, . . . - . ---c - , ,,
- CHANGE
- NUMBER EE DESCRIPTION Itlhati As.s-9 (
5.5 alled in a s chan istrati contrfs _ 'nl of t TS. s g ^res inp i nge t echni _ requir nt y 4
- 12 05 TR 1 Revised Surveillance Requirement to allow for increased '
flexibility in using an actual or simulated actuation i signal. Identification of the specific signal is moved to , the Bases. ! 12-06 A This requirement would demonstrate the operability of each , train of the [ Emergency Exhaust System] (including i maintaining negative pressure in the building). This is , consistent with current practice. This change does not i result in a change to technical requirements and is consistent with NUREG 1431. Rev. -1. i _ _ _ _ f-%3. M3.7.-3 l { 12 07 LS 25 The propos ~ ange woul remove "ST RED TEST IS" l iusatT 3A-%. , from t day SR. is repres no change the i _ fr ncy of te ng since t S definiti of STAGGERED
- BASIS d have testing e of the -
l [Emer Exhaust Sys ] trains ev 31 days tg j' _reguire ea rain to be ted every days. s KAu.-eel l 12 08 LS-16 The proposed change would allow the 18} month testing of i the [ Emergency Exhaust System's] ablity to maintain the i required pressure differential between the building and the outside atmosphere to be performed on a STAGGERED TEST BASIS. 12 09 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 12 10 LS 9 The "within 31 days after removal" requirement for completion of laboratory analyses is deleted. This requirement is not contained in the ITS nor is it contained in the regulatory guide or ANSI standards. 12 11 A The SR to measure r argency Exhaust System] flow rate is not identified as a separate SR in the ITS because it is ! verified during the other in place filter tests (see ITS I 5.5.11 a. and b.). This change does not result in a change to technical requirements, i WCGS-Description of Changes to CTS 3N.9 9 S/15/97
, _ . . . .- . . - - . ~.-- . .. __.-. _ _ . . - _ . . . _ _ -
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 5.5-10 APPLICABILITY: WC, CA l REQUEST: ITS 5.5.11.b (Callaway and Wolf Creek) l Comment: The smooth copy of the ITS still has the [] around the plant specific bypass value FLOG RESPONSE: The smooth copy of the ITS has been marked to delete the brackets ([]) around the plant specific bypass value. A final review of the smooth ITS and ITS Bases is planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases ATTACHED PAGES: None I t l l l i
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-12 APPLICABILITY: WC ! l REQUEST: ITS 5.5.11 and CTS 4.7.6.c.2 (Wolf Creek) l Comment: The value of relative humidity is 70% in the ITS,78% in the CTS markup, and 70% in the l CTS. Is it correct to assume the CTS markup value is wrong? FLOG RESPONSE: The 78% value for relative humidity in CTS 4.7.6.c.2 and CTS 4.7.6.d in the license amendment request is a typographical error. The correct value is 70%. The CTS mark-ups have been revised to specify the correct value. ATTACHED PAGES: Att chment 13, CTS 3/4.7 - ITS 3.7 Encl. 2 7-15 1
l 1 PLANT SYSTEMS
- SURVEILLANCE REQUIREMENTS (Continued) ._ . _ _ _ _.---.-- - .---_. - ..--- . . ._.- .
(New) Perform required CREVS filter testing in accordance with the Ventilation Filter Testing Program (VFTP); ggg]
- c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following gn}} -
painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the Control Room Emergency Ventilation System P{ S ASM
^
*M i
satisfies the in-place penetrabon and bypass leakage testing acceptance criteria; of less than 1% for HEPA filters and 0.05% for charcoal adsorbers and uses the test procedure guidance in Regulatory Posibons C.5.a. C.5.c, and C.5.d of Regulatory Guide 1.52 Revision 2. March 1978, and the system flow rate is lQ s.5 l l 2000 cfm 110% for the Filtrabon System andf2200minf'h&O54df f*jMSM}l} M: - -' -- -u i.+; ;7750 cfm 110%y ,fr.,opy3the Pressunzation System filter adsorber unit;
- 2) Venfying, . _ . ?? dr; :":: nz: ::, that a laboratory T}M 0MIK
* *%3d, analysis of a representative carbon sample obtained in accordance N 1 with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testi criteria Q* ]
of ASTM D3803-1989 when tested at 30* C and relative g,g.a g humidity, for a methyliodide penetration of less than ~10*h 2%; and
- 3) Verifying system flow rate of 2000 cfm 110% at greater than F 10 l'f4 '"'l
-- '" "-*d '
7 = anal to 8.1 Ir chen W.G. Idirty filter) for the Fir. ration 0b ystem andG2OlemwSPAiSt greater than or equal to 3.6 % s.5 T { ['le
. inches W.G. { dirty filter) for thetelessunzaticoevawmJnim - ^ ~1$-LG
+ ~ ^ ~$ "
731>tmp410Hrfomg1brougMh_eJPressurization System filter adsorber unit durir g system operation when tested in accordance with ANSI N510-1980.
- d. After every 720 hours of charcoal adsorber operation by verifying T 10 23-LS-13 P
'^". 21 dr;: ":: nz:::: that a laboratory analysis of a represen- " --" ^ 4 ^'
tative carbon sample obtained in accordance with Regulatory Position ' '" 10 08-A" ",
-~~ - - " J~
C.6.b of Regulatory Guide 1.52, Revision 2. March 1978, meets the labor testing criteria of ASTM D3803-1989 when tested at 30'C and lative humidity, for a methyl lodide penetration of than 2%; - (~M*h y.Q s,S 2. \
- e. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA l filters and charcoal adsorber banks is less than 6.6 inches M**10 ~ "" 08!AK l Water Gauge while operating the system at a flow rate of 2000 cfm )
10% for the Filtration System and less than 3.6 inches Water Gauge while operating the system at a flow rate of 750 cfm 110% for the Pressurization System filter adsorber unit
- 2) Verifying that onfan actual or simulated actuatioit: C020! Poem y .,_.= g._.::.er T .9::: r :r:- ^' t;!::t signal, the
#NH010-TR
*Q*
system automaticallykctuates?:' 2".:t:: ht: : r-Mr"' " med? Of 0;0'f 0n
"' ^r thr"-t St M5P^ 90- 2nd Ot;:-- ' :100- 5 b: t l
I WCLF CREEK - UNIT 1 3/4715 Amendment No. 22-102 Mark-up of CTS 3N.7 5/158 7
l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 5 6-1 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 5.6.5 a.7&8, Changes 03-14&15 M Comment: it is true that the additions would make the COLR more restrictive however, the removal of the specific values from the TS is a less restrictive change that needs to be justified. Provide justification. l FLOG RESPONSE: DOC-03-14-M describes the addition of the SHUTDOWN MARGIN (SDM) limits and the Moderator Temperature Coefficient (MTC) limits to the Administrative Program description of the CORE OPERATING LIMITS REPORT (COLR). As stated, this change is more restrictive to the COLR. The change for moving the actuallimits from the technical specifications to the licensee controlled COLR are addressed and justified by DOC 01-01-LG (SDM) found in Section 3.1 (not applicable to CPSES) and DOC 03 LG (MTC) found in Section 3.1 (applicable to DCPP only). DOC-03-15-M, in a similar way, adds the Refueling Boron Concentration limits to the Administrative Program description of the COLR. The change moving these limits to the l licensee controlled COLR is addressed and justified by DOC 01-02-LG found in Section 3.9. ATTACHED PAGES: None l i l l
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.7-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 5.7.2 and Difference 5.7-2 Comment: TSTF-167 has been rejected by the NRC. Use current ITS. FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. This traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the April 9,1997 letter from C. Grimes (NRC) to J. Davis (NEI), with i some exceptions. The FLOG submittals have been revised to incorporate TSTF-258 and encompass the NRC cornments of 6/11/98. Additional technical changes made to Section 5.7 are identified and justified (See JFD 5.7-2 which revises ITS 5.7.2e consistent with CTS 6.12 and JFD 5.7-4 which revises ITS 5.7.2f consistent with CTS 6.12). The latest industry status on TSTF-258 is that the NRC has requested changes to Section 5.7, High Radiation Area. l ATTACHED PAGES: Sse markups associated with Comment Number Q 5.2-1. 4
l ADDITIONAL INFORMATION COVER SHEET I- ADDITIONAL INFORMATION NO: CA 5.0-003 APPLICABILITY: CA, WC, DC REQUEST: ITS 5.5.10, is revised to delete the words: "and low pressure turbine disc stress corrosion cracking". This requirement is not part of the Secondary Water Chemistry Program described in
- CTS 6.8.4.c.
l ATTACHED PAGES: Encl. 5A 5.0-21 Encl.6A 4 Encl. 6B 4 I
Programs and Manuals 5.5 l 5.5 Programs and Manuals 5.5.10 Secondarv Water Chemistry Proaram l This program provides centrols for_ monitoring secondary water chemistr ; to inhibit SG tube degradation :Ed'. ,;ti:Qregurbtffe drse strys] s.s-r1 l cpFrosp SPacMnjp The program shall include: Ic A 5.o-cc3\
- a. Identification of a sampling schedule for the critical variables .
and control points for these variables: l
- b. Identification of the procedures used to measure the values of the l critical variables: I
- c. Identification of process sampling points, which shall include ;
monitoring the discharge of the condensate pumps for evidence of ; condenser in leakage:
- d. Procedures for the recording and management of data;
- e. Procedures defining corrective actions for all off control point l chemistry conditions: and I
- f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.
5.5.11 Ventilation Filter Testina Proaram OFTP) ,#. r BiPSiM A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1152._ Revision 2, 'ggg and in accordance with the guidance specifi.ed below. -
- a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0-05 1% when tested in ' c 0"O l accordance with Regulatory Guide 1.52. Revision 2 rend-ASI E :_ B-PS ' 1 510 1000 at the system flowrate specified below i 10%.
y (continued) WCGS-Mark-up ofNUREG-1431-ITS S.5 5.0 21 $/15/97
~
s.s -e U GA ff l ' CD A O N
- 0"O O*"P ***" I'MM' sun]aec, s.o.od}
HANGE ' NUMBER J.USTIFICATION 4s.ii iusent (A- 4Mo 9.5-7 ) 5.5 15 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). C. L.s - u. 'thSE RT M_-D) G S.1- I \ 5.6 1 Not applicable to WCGS. See Conversion Comparison Table (Enclosure s.s C-F ' tN h.8~T(=A-44) ' 5.6 2 This change deletes the Emergency Diesel Generator Report to reflect the recomendations of GL 94 01, " Removal of Accelerated Testing and ) Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. j i 5.6 3 This change revises the report date in Section 5.6.2, " Annual ) Radiological Environmental Operating Report," to be consistent with l current TS . 5.6 4 This change revises Sections 5.6.1 and 5.6.3 " Occupational Radiation l Exposure Report" and " Radioactive Effluent Release Report," respectively, per NRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From Christopher I. Grimes to Owners Groups Chairs). This change is consistent with traveler TSTF-152.
)
5.6 5
~ O[ ] PORV lift settings are referenced in the PTLR section r5TF-2 pe
- 64. -co _ lusser G A -4 bM os;.z-i 1 ----
5.7 1 (This~ ang evise High ia n AJ # a to gi orporptf chya p!s7 f co st with 0C 0. 1)./Speciftfally, distanc(s fro (the 5.% ; tadia on so ce are ot . t ys,se.rt A .4 c. j j 5.7 2 /Thi cha ergses" naut riz t "in vert t" nteH di on fa se on ref ct e NR s iti n a sta ed 'n /
/ Re a Gui 8.38 Sec n1 reg din phy cal arr er for/High 1 adia n Ar s. T si cons tent ith av er T F 67.
5.7 3 (Not applicable to WCGS. See Conversion Comparison Table (Enclosure ___ i usEcr s.A - A cL. 68). l . __ l 4 Ch&y Q S.2- 0 WCGS-Differencesfrom NUREG-1431 -ITS S.O 4 S/lSB7
INSERT 6A-4a 0 5.2-1 5.5 16 The Radioactive Effluents Controls Program is revised to include clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. These statements of applicability clarify the allowance for surveillance frequency extensions and allowance to perform missed serve 111ances. Generic Letter 89-01, " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications and the Relocation of Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents Control Program in the Administrative Controls Section of the ! Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 89-
- 01. Since this change adopts previous CTS requirements, it is considered a change of presentation method only. This change is consistent with TSTF-258.
INSERT 6A-4b 0 5.2-1 5.6-6 The ITS requirement to provide documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694, "THI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states: t
" Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report." NRC l
Generic Letter 97-02, " Revised Contents of the Monthly Operating Report" requests the submittal of less information in the monthly operating report. The generic letter identifies what j needs to be reported to support the NRC Performance Indicator ! Program, and availability and capacity statistics. The generic letter does not specifically identify the need to report l challenges to the pressurizer safety and relief valves. This change is consistent with TSTF-258.
. INSERT 6A-4c 0 5.2-1 Section 5.7 is revised in accordance with 10 CFR 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. These changes are consistent with the draf t Generic Letter (93-XX) on proposed changes to l STS NUREGs based on the new 10 CFR 20 and the letter from C. Grimes, NRC, to l J. Davis, NEI dated April 9,1997. This change is consistent with TSTF-258
- and encompasses the NRC comments on 6/11/98. Additional technical changes made to Section 5.7 are identified and justified. 4 t I i
l INSERT 6A-4d 0 5.2-1 l ITS 5.7.2.e is revised consistent with CTS 6.12 that allows any individual or group of individuals to enter a high-high radiation area (dose rates greater than 1.0 rem / hour at 30 cm) accompanied by an individual qualified in radiation protection procedures with a radiation dose rate monitoring device. The qualified individual is responsible for providing positive control and shall perform periodic radiation surveillances at the frequency specified in the RWP. The CTS requirements allow the qualified individual to enter a , locked high radiation area with plant workers without first having to enter i the area to determine dose rates and then exit the area to provide dose rate information to the plant workers and then reenter the area. This flexibility ; is in keeping with the "As Low As Reasonably Achievable" principle while ' maintaining appropriate radiation worker practices. INSERT 6A-4e 0 5.2-1 l L ITS 5.7.2.f is revised consistent with CTS 6.12 to delete the phrase "that is controlled as a high radiation area". The proposed change would preclude having to post an area around the high-high radiation area as a high radiation i area when the area may not meet the definition of a high radiation area.
]
INSERT 6A-4f CA 5.0-003 ITS Section 5.5.10 is being revised consistent with CTS 8.8.4.c. The proposed change deletes the phrase "and low pressure turbine disc stress corrosion cracking" from the ITS to be consistent with the practices of the CTS which do not have this requirement for the Secondary Water Chemistry Program. INSERT 6A-4g 0 5.5-7 5.5-19 CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety-related ventilation system filter adsorber units include the requirement to measure flow rates within specified values. while imposing an artificial differential pressure, during system operation. when tested in accordance with ANSI N510-1980. This flow rate testing is to be performed at least once per 18 months, after any structural maintenance on the HEPA filter or charcoal adsorber housings. or following painting. fire, or chemical release in any ventilation zone communicating with the system.
. Therefore. the CTS surveillance requirements are incorporated into the ITS.
i r l
L o ! CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 4 of 5 SECTION 5.0 ' i DIFFERENCE FROM NUREG-1431' APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY ; NutBER DESCRIPTION i No No 5.5-12 The referenced frequencis for the tests listed in the Yes No Ventilation Filter Testing Program (VFTP) were evaluated i as part of the 24 month fuel cycle program for DCPP (see ; LAR %-09) Revises Radioactive Effluent Controls Program dose Yes Yes Yes Yes i 5.5-13 projections to meet original intent of TS prior to implementation of GL 89-01. (WOG-72) l t 5.5-14 Section 5.5.7 is being revsied consistent with traveler Yes _ Yes Yes Yes I
% [and faMg iMUle@es>'fubgMGd Defenberv g g Q t >1C to(. -{QS.5-2.\
(PD ./ . The proposed changes to Section 5.5.7 provide gg 3 q g ; an exception to the examination requirements in Regulatory Guide 1.14. Revision 1. " Reactor Coolant Puup Flywheel .l Integrity." No No Yes 5.5-15 This change provides a tin't interval of within 31 days No after removal in which a laboratory test of a semple obtained from the charcoal adsorber must be tested. This change is consistent with Callaway CTS. t Yes. LAR 94-14 No. Wolf Creek CTS No. Callaway CTS 5.6-1 Revises Section 5.6.4. "61onthly Operating Report." to No. DCPP CTS consistent with consistent with consistent with ! reflect a revised submittal date. NUREG-1431. NtREG-1431. ' NUREG-1431. No. Not in CTS No. Not in CTS. No. Not in CTS. , 5.6-2 Deletes the EDG Report t) ref1?ct the reconnendations of Yes equir 'I Di ---fG s.6-2_ } ) Jiat y 31. . TSTF-3 7, Rev 1. < Yes. Consistent Yes. See LA 42/28. Yes Yes 5.6-3 Revises report datu in ITS 5.6.2. " Annual Radiological Environmental @erating Report" to be consistent with with CTS and LA j r current *5. 78/77. (s.s-L' msee:r se; 4plTs.c i ) s.s-th[sse:r e,sf-4r) gos.s 7 l ~ . ImLT 6r[- 4bM CA S.O-OQ1 } t uw-r r,e> _4 c_yg oc s,o . co.3 {
~
(c.s - i$ 5/15/97 l WrG%Gmvenian Gumtwriwn Tame - ITS 5.0 !
INSERT 6B-4a 0 5.2-1 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION
- The Radioactive Effluents Controls Program Yes Yes Yes Yes 5.5-16 is revised to include clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities.
INSERT 6B-4b CA 5.0-003 . TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY I NUMBER DESCRIPTION Yes Yes 5.5-17 This change deletes the phrase "and low pressure Yes No turbine disc stress corrosion cracking" from ITS 5.5.10 to make the program consistent with CTS 6.8.4.c. i DC 5.0-003 INSERT 6B-4c , TECH SPEC CHANGE l APPLICABILITY : DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION Yes No No No 5.5-18 Revises DCPP Sections 5.5.9 and 5.6.10 to reflect License Amendment 124/122. dated March ,
- 12. 1998, which allows implementation of steam !
generator tube voltage based on repair criteria for ODSCC indications at tube to tube support plate intersections.
INSERT 6B-4d 0 5.5-7 TECH SPEC CHANGE APPLICADILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK UOLF CREEK CALLAWAY No No Yes No 5.5-19 Wolf Creek CTS surveillance requirements 4.7.6.3c.3) and 4.9.13b.3) for safety-related ventilation system filter adsorber units include the requirement to measure flow rates within specified values. while imposing an artificial dif ferential pressure. during system operation, when tested in accordance with ANSI N510-1980. The CTS surveillance requirements are incorporated into the ITS.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 5.0-003 APPLICABILITY: CA, CP, DC, WC REQUEST: "lTS 5.6.6, " Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)", was revised to incorporate changes based upon WOG-67. WOG-67 has l been approved by the TSTF and is designated as TSTF-233. This traveler has been submitted I to the NRC and the latest traveler reports indicate that TSTF-233 has been approved by the NRC. The attached pages reflect changes associated with WOG-67 being designated as TSTF-233." ATTACHED PAGES: Encl. 5A Traveler Status page Encl. 6A 4 Encl. 68 5 i 1 ( L
f INDUSTRY TRAVELERS APPLICABLE TO SECTION 5.0 TRAVELER # STATUS DIFFERENCE # COMMENTS l TSTF-9, Rev.1 Incorporated NRC approved. TSTF-37, Rev.1 Incorporated 5.6-2 DCPP only. _ TSTF-52 Incorporated 5.5-4 ( qb$hq 3,u.g l TSTF-65 @ Incorporated NA MNRC approv y,,g a.o s] 8 s z-s l cit. w t sd , _ j TSTF-106, Rev.1 Not Incorporated NA Retain CTS. j TSTF-118 Incorporated 5.5-8 %ep e@lTe.r.o-ooc.)
@ T 611 V yafluc3rporate3-' ,, MA /Retaja CJ8} l re. c.o -ooc. J l TSTF-120@ Not Incorporated NA Retain CTS Fra s.o-ox. 1 l
Wi Ip6er6edeV_ / Sa-E !o s.2-0 TSTF-152 Incorporated 5.6-4 %Qera] 4 ra r.o-co6\ (T.s @ As 9 ./Jgotp(r'ated / / 5:7A") f @ 6 Z-83 h.08%WRe@ Incorporated 5.6-5 hhoQ4 TR s.o-aaj WOG-72 Incorporated 5.5-13 Incorporated 5.5-14 !* * \ l beppeed Trsteler Incorporated s.2-2, 5.5-1,s.23, oupAc "T'Tp - 158 ps.2.-i }
*21,>lQ-2,S.C-'%
., e 4.
i S/158 7
3 ,5 _te, O pyle cau -la (X&s, 6ea. (onw.ruf n Compensen~TaRc.(Ent. % s.o.co HANGE ' NUMBER JUSTIFICATION @i sus nr a-4M 4> s.s-7 ) l l l 5.5 15 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). Cs.s-u._ 'lM6E Rf G=A -D) G S.1- L \ 5.6 1 Not applicabie to WCGS. See Conversion Comparison Table (Enclosure s.s 67 iN E.Wr(oA-44) 5.6 2 This change deletes the Emergency Diesel Generator Report to reflect the recomendations of GL 94-01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. 5.6 3 This change revises the report date in Section 5.6.2, " Annual l Radiological Environmental Operating Report," to be consistent with l current TS . 5.6 4 This change revises Sections 5.6.1 and 5.6.3 " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report," respectively, per NRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From , Christopher I. Grimes to Owners Groups Chairs). This change is consistent with traveler TSTF 152. 5.6 5 6[ ] PORV lift settings are referenced in the PTLR section STF-2 W' Cs. c. - co lusser c, A -4 Ebi os.z-i 1 --- 5.7-1 (Tfn s ~ ang evise High ia ' n AJ# a to p(orporptf chgasf_ 7 1]./Specif4fally, distancfs fronithe co st with 0 CF 0. tadia on so ce are ot . luSELTfoA -4 c. j
~
5.7-2 by'hIer ses " naut riz t "in vert t" nteH JWdia on a se on ref ct e NR s iti n a sta ed 'n / Re ato Gui 8.38 Sec n1 reg din phy cal arr er forAligh F- 67. (Notadiatapplicable on Ar s. T si cons tent ith I av er T _ l usEci AA- Ack. 5.7-3 to WCGS. See Conversion Comparison Table (Enclosure 6B). @+ pp +e-y a s.2- d WCGS-Di(ferencesfrom NUREG-1431 - ITS S.0 4 S/1S/97
CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 5 of 5 , SECTION 5.0 DIFFERENCE FROM NUREG-1431 APPLICABILITY NLHBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.6-4 Revises Sections 5.6.1 and 5.6.3. " Occupational Radiation Yes Yes Yes Yes Exposure Report" and Radioactive Effluent Release Report." respectively, per NRC letter dated July 28, 1995. " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (from Christopher I. Grimes to Owners Groups Chairs). This change is consistent with traveler TSTF-152. 5.6 5 DCPP LTOP a' ..; and PORY lift settings area efer Yes Yes Yes Yes PTLR section perWMsTP-233(nced -
)- in - j iR S.0-603 }
st rc,B R Q S .2. t] 5.7-2 fiIinges "unauthor " to "inadvert t" in the High , W W W -Ves- pg 2_q Radiation Are ection to ref1 the s position as
! stated i 8.3.8. Section reg ing physt6al barr s for High Radia n Area This clienge is (sistentwithtrave6rTSTF-
- 7. l*lSERT 6B-Scy 5.7-3 This change deletes the phrase "or that cannot be No No No Yes contiauously guarded" from the ITS for Callaway to make -
them consistent with the CTS. iklsEAT Qs.t.-1] 4 twscaer 66 - 6 {Qs.2.-l { WCGS-Conversion Comparison Table-ITS5.0 S/25/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 5.0-006 APPLICABILITY: CA, CP, DC, WC REQUEST: Revise the Traveler Status Sheet to reflect the latest status and revisions of the follov:lng travelers: TSTF-118 - NRC Approved TSTF-119 - NRC Rejected , TSTF-120, Rev.1 l TSTF-152 - NRC Approved ' ATTACHED PAGES: Encl. 5A Traveler Status page 4 l
l l INDUSTRY TRAVELERS APPLICABLE TO SECTION 5.0 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated NRC approved. TSTF-37, Rev.1 Incorporated 5.6-2 DCPP only. TSTF-52 Incorporated 5.5-4 ( q3 ha 3,u .c j TSTF-65 @ Incorporated NA hNRC approved } yg,g...oes )
@ S 2 'l Ct@veles-(1iitord
)
TSTF-106, Rev.1 Not Incorporated NA Retain CTS. TSTF-118 Incorporated 5.5-8
%eer@lTc.c.o-coc.I
@l1V jief'Inc3rp6 rated - '
#A /3etaja CJ}8 l ra. c.o -oor. j TSTF-120@ Not Incorporated NA Retain CTS Prx s.o-ow. 1 WW Ip6er6erdteV_ / 52 {QG.1-L)
TSTF-152 Incorporated 5.6-4 gQ 4 raco-ooQ (} 8'f h N /Wotpfrasetf / 7 5:7,r) i 4 S- -d hkM7 Incorporated 5.6-5 (p[c.ipproul)@R s.o-at3] WOG-72 Incorporated 5.5-13 Incorporated 5.5-14 I* ** ~
- I f.sepped Tritteler Incorporated s a-2, 5.5-1, s.z .s, sp, Wpp oupAc)i6n pg,z.,1 g rsrs:- 2 56 .
,,3.7,2, s g-e ftenpf4 .
SAS/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 5.0-ED APPLICABILITY: WC i l l REQUEST: 1) The electronic CTS page 6-18a did r:ot have the correct text for item b. under CTS 6.8.4.g. This page has been marked to correctly reflect the CTS. I
- 2) DOC 3-07 in Enclosure 3B should be identified as 3-07-A.
1
- 3) In ITS 5.5.11c the "s" is deleted consistent with CTS SR 4.7.6c.2) and ASTM D3803-1979. l ATTACHED PAGES:
l
- Encl. 2 6-18a i Encl. 3B 6 Encl. 5A 5.0-22 I
l l l
- . - - - . . . - _ ~ - . --- .. .
l l l ADMINISTRATIVE CONTROLS l PROCEDURES AND PROGRAMS (Continued)
- 3. a kinematic viscosity within limits for ASTM 2D fuel oil,
- 4. a water and sediment content within the limits for ASTM 2D fuel oil;
' 02-16-A '
- b. Other properties for ASTM 2D fuel oil are 'c"' "-it(analyzed lwithin (days
- following sampling and addition of new fuel oil to storage tank ; and 02-19-LS-2 I
- c. Total particulate concentration of the stored fuel oil is < 10 iter when tested every 31 days based on ASTM D2276, Method A.
w h e m h..h e.d.tra w A o n. 02-12-LG
' Eme= ne;OH 'C=:=te od .
T' omr:r gmpg . lac. S.6-ED3 a, :m=;:=y ere!;=:=tr =!! !!?; ;m;=m 'h:t et 9 = th: =; !===:: Ond ;r d:"n : 'r ;m;.;;n y frr' ;r =tr - "M"?;,27 "^'"?;,2nd ==!!:ing. The ; ; Tem th !'2^^'U I th0 ff Jf.;:
- 2. Ern;: . / fr^';= retr =!:d!"?; p^d - nr ;- '- (te; t r"f:'t;) bird up n 'h: :!f:n 5' drut r;!n; crrrren' Tr;;! r"f"!!!;rr' m nd:t;i: i
- 1;'!:'. d t' rr;'. m:nn:t; m:!'.02 th t r: brrd up . 'hrr tre:d ?-
A---C M ad L11 IRE A SF O'F nn
.rr------'- -- - - ' - - - - - -a
- b. r r !: Oru= tt:":d met errr nd; !: Of Omr;rr; fer';rr:tr
' "r r !:;;ix:d 2nd :":f= cr=d= cfr= = iden :- e;en: t ' "ure,
- c. 't;':mrtfen Of en :mr;=r; fer!;=:=tr p=r:n= m:!nten== ;=;=m
*c' b
- d. rM;t; ef Omr;x;; frc' grr:tr 2" "d"t; =d ;:t= = pr:m:':=
!::=== th: t r;:t r"d!!?;i: met e rn:d:d.
I. Containment Leakace Rate Testino Procram A orogram shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regu;atory Guide 1.163, Performance-Based Containment Leak-Test Frogram," dated September 1995. The peak calculated containment intemal pressure for the design basis loss of coolant accident P., is 48 psig. WOLF CREEK UNIT 1 6-18a Amendment No. SG-GA.101 Mark up ofCTS6.0 SMSM7
CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 6 of 8 TECH SPEC CHANGE APPLICABILITY . NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CAlLAWAY 03-02 The requirement to submit a startup report is deleted from Yes No. Deleted from Yes Yes A the C15. This report required no staff approval and was CTS per Amendment submitted after the fact and, is therefore, not required to 50/36 ; ensure safe plant operation. The approved 10 CFR 50. Appendix B. QA Plan, and FSAR startup testing program provides assurance that the affected activities are adequately performed and that appropriate corrective actions, if required, are taken. 03-03 Revises the annual report section to reflect the new 10 CFR Yes Yes. Except the Yes Yes A Part 20 requirements and associated recomended changes Part 20 noted in NRC letter dated July 28. 1995. " Changes to requirements were Technical Specifications Resulting from 10 CFR 20 and re eved from the TS 50.36a Changes." (From Christopher I. Grimes to Owners in A Endment 50/36 Groups Chairs) - TSTF-152. , 03-04 The requirement to report specific activity limit Yes Yes Yes Yes A violations is deleted consistent with NUREG-1431. Serious degradation of a fission product barrier, among other more serious events are required to be reported by 10 CFR 50.73. This change is adninistrative in that it only affects reports and do not affect plant operations. 03-05 The Annual Radiological Environmental Operating Report No. DCPP report Yes No. WCNOC report No. Callaway A including submittal date is rr. vised. dates to remain as dates to remain as report dates to in CTS. *n CTS.
. remain as in CTS Yes Yes Yes Yes 03-06 CTS [6.9.1.7] " Annual Radioactive Effluent Release A Report" and CTS [6.14.c] is revised consistent with NUREG-1431. Rev.1. to delete the term " Annual" and modify the submittal date.
03 07 Yes Yes Yes Yes[I4C So-ED ]
@ CTS [6.9.1.6]. " Annual Radiological Environmental Operating Report" is revised to include specific details concerning the contents of the report.
WCGS-Conversion Comparison Table- CTS 3N.0 S/15/97
Programs and Manuals 5.5 l 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testina Procram (VFTP) (continued) ESF Ventilation System Flowrate Control Room Emergency Ventilation System Filtration 200.0_cfm .. i l Control} Room!Emergenc[Ventilafion' System _- Pressurization 750"cfm .dil8?PSW_8
'^
Auxiliary /FuerBufTdi_ngT.mergency T.xhaust _
~ 6500 cfm
- b. Demonstrate for each of the ESF systems that an inplace test of the charcoal absorber shows a penetration and system bypass
< [0.05]% when tested in accordance with Regulatory!Guide 1.52, g' ggp3;" '
Revision 2. und AS".E N5101000 at the system flowrate specified below Et 10%3
~ ESF Ventilation System Flowrate_
kn+ro=l Room Emergeny Ventjbhpn gtaa - Fdtvahm 2cco cbf N' 8- ~0* 6 l Control . Room Emergen.gy Ventilation System T Fressurization 750 cfai MEN cAuxiliary/fueliBui,1dingiEmeMyjExhaust kki5n l
- c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1~.52, Revision.2, shows the methyl iodide pgg;p3g penetration less than the value specified below when tested in accordance with ASTM D3803 1989 atatemperatureof@30*Cand , 58%
- greater than or equal to the relative humidity specified below3 Mwcs.o-eol ESF Ventilation SyJst __ Penetration RH l (( Si+nNm / Preesh<f=a.WB ; wc. s.o-oor. l l Control Room Emergency Ventilation. System i 2% 70% pgg3psy .
Auxiliary / Fuel ~ Building Emergency Exhaust 2% 70_t
.;vi;w;r;
Netc. All;w el; g netr; tion - [1000 r.;thyl iodid; offici;ncy for ch;r;;;l ci;dited in ~; toff ;;f;ty CVol.; tion]/ (';;f;ty f;cter'. S;f;ty f;;ter[5] for ';y:;tc.a with h;; tor;.
- [7] for ';y:;t;:;; without h;;tcr'i.
I 9 l (continued) WCGS-Mark-up ofNUREG-1DI-ITS S.0 5.0-22 S/1S/97 l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 5.0-001 APPLICABILITY: WC REQUEST: Amendment No.110 was issued on September 22,1997. This amendment modified CTS 5.3.1, l
" Fuel Assemblies" and CTS 6.9.1.9, " CORE OPERATING LIMITS REPORT," to add ZlRLO as l
fuel material and the use of limited zirconium alloy filler rods in place of fuel rods. As identified in Attachment 3 to the conversion application, the amendment request was incorporated into the conversion application. Subsequent to the submittal of the conversion application, the ZlRLO license amendment request was revised at the NRC request to include the reference to WCAP-12610 in CTS 6.9.1.9. This licensee identified item reflects in the conversion application the approval of Amendment No.110 to the Wolf Creek license. Additionally, the numbering in ITS 5.6.5 is corrected. ATTACHED PAGES: Encl. 2 6-21b Encl. 5A 5.0-32 l l [
., ~ ~, . . . . . - _. - - -. .. . _ . . . . _ . - . _ _ . . _ . . . . . . . . . ~ _
ADMINISTRATIVE CONTROLS I CORE OPERATING LIM!TS REPORT (COLR)(Continued) w _i _ _ r- V
- _ _ _ . e._._L _ . . ,_,m-.o.
- __u...,r.k.._._.._._,e__...
. . . . . . . .. . -. _ a w.e _,__ .a .:_ - . _
l
,, a. 4. .k mas.._e__n__.._c.__.._._._..__.__s,
~ ._.-_ . . _ . . . .. . _
L . WSEST' 2, - 2 l3 ' taJC G.0 -o o I L The core %.ung umits snan be determined so that all app icable limits (e.g., fuel thermal-hydraulic limits, core thermal-hydraulic limits, j ECCS limits, nuclear limits such as shutdown margin, and transient and l accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or l supplements thereto, shall be provided upon issuance, for each reload cycle, i to the NRC. Der- .: : C;n^ :! D::h "'". r;!:: t: the " ;': :' ^.d " .! .._: r end .
"::S::: ' :;: ^:- 03-08-A 2_n SAEGIAIAEAGRT4 43 n A~
e___:_s-___2_
. a. =n3w ..e _ _._ _- - --._
_ L _ se L. _ . ' * * * - J a
- w. w.. w. ---. . . - _ g. ak.
- i. w . _ _ ? _ _ AJ m.wgw. .
_ t._.....-."...".g,_..,&k.. - .. .w
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r RCS Pressure and Temperature Limits Report (PTLR)'i new 03-13-M see insert 21 .w.rL PAM Report (new see insert 22 j l I i l t' WOLF CREEK - UNIT 1 6-21b Amendment No. 64, HO d UC S 0-08 '] Mark up of CTS 6.0 S/1587
. ..__ ~. . _ _. .._ - . - . _ . _ . _ . . _ - . _ . . _ , _ . _ _ _ . _ . . , _ . _ _ _ . _ _ _ _ _ _ _ _ _ . - . .
! \ i INSERT 2-21 WC 5.0-001
- j. NRC Safety Evaluation Reports dated-July 1, 1991, " Acceptance for ,
Referencing of Topical Report WCAP-12610, ' VANTAGE + Fuel. Assembly ! i- Reference Core Report' (TAC NO. 77258)," and September 15, 1994,. '
" Acceptance for Referencing of _ Topical Report WCAP-12610, Appendix B. ;
. Addendum 1. ' Extended Burnup Fuel Design Methodology and ZIRLO Fuel .
- Performance Models' (TAC No, M86416)" (WCAP-12610-P-A) !
)
(M0th0dO'.0gy f0F Speci#iCatiOr 2.2,2 - Heat E'.ur. 40t Char".01 F20t0F - S- r1 -A ~ f !: Fotl4
~
l-I l I 1 l l l' 1 i l
, y , w , -. -
Reporting Requirements 5.6 5.6 Reporting Requirements 1.; NRCf Safety _ Eval uation _ Report @ted 70ctoberi29fl992,'ifgr l theTCortfThermal?Hydraillic[AnalysisiWethodologyffor;the Wol.flCteek; Generating (Station;"3Eri90iV140 ZEE 92f0103) . _
- 2. ECiSafety, Evaluation Reportidat_edRanuaryJ1LJ1989,1for the;3cceptance for Referencing;ofC1censingJopjcal Report ~WCAP;11397. RevisedfThermal .1)esi~gnhedure.?
- 3. EC_ Safety; Evaluation 3eportidated: September 330M93ifor the Hransient" Analysis Hethohlogygthe Wolf; Creek l Genergting;StationT(ET'-91f0076[lTF9230142[WN L93!0010, WN93i0028).
! 4. NRC.; Safety _Evaluat_1on Report; dated.NoyenberJ26,i1993, "Acc~eptance;for Referencing!oflReyvj; sed Version ~of LicensingJopical ReportiWCAPi10216iPJAERe1axation.of Const.adt:Ax1all Offset ControTOFi!Surynill'anceTechnical Specjfication" (TAC _No. N88206L
- 5. NRC ; Safety; Eval uation } ReportTdatedLHarch30f]993 Rforithe "ReloadlSafety EyaluationMhodology forithe WolfiCreek Generating 7 Station?l(EE92;0032i;EE9310017) .
- 6. EC Safety _ Evaluation Report" dated lHarch'30E 1993;1for the
" Revision;tgiTechnical Specification"for.; Cycle 7"'(NA 92-l 0073, NA 93-0013, NA'93 0054).
wcs.e-ee 4 NRC Safety Evaluation. Report"Aa.ted November l13,1986, mr "The 1981 Version of the Westinghouse 1ECCS;Eyaluation l Nodel Using the BASH Code"l(WCAP 10266-P A, Rev.f 2). NRC _ Safety, Evaluation... Report _ dated May 17,,1988, Ob7. " Acceptance for Referencing.of: Westinghouse Topical. Report WCAPil1596,:.. Qualification, of.ithe Phoenix-P/ANC' N.uclear DesignfSystem for Pressurized ~Watet Reactor _ Cores." [WC S.O-oo Ij 9E EC_SafetyiEvaluation Report;datedlunel23,3986.l
" Acceptance ~;for Referencing lof Topfca13eport WCAP 10%5 P and WCAP110966 NP-ANC: A' Westinghouse' Advanced: Nodal i _ _ _ W ' ? ~ _
l . Nsr.Kf SA 3 [QC LO~OO I \ i _ I i (continued) WCGS-Mark-up ofNUREG-1431-1TS 5.0 5.0 32 S/15/97 l _ _ _ . __ _ _ _ . _ _ .
r-l ! INSERT SA-32 WC 5,0-001 L l 10. NRC Safety Evaluation Reports dated July 1.1991, " Acceptance for Referencing of Topical Report WCAP-12610, ' VANTAGE + Fuel Assembly Reference Core Report' (TAC N0. 77258)," and September 15, 1994,
- " Acceptance for Referencing of Topical Report WCAP-12610 Appendix B, I
Addendum 1, ' Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC No. M86416)" (WCAP-12610-P-A) l. l l l l r
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 5.0-002 APPLICABILITY: WC REQUEST: The word " Manuals " is misspelled in the header on page 5.0-7 through 5.0-26. The editorial change has been identified for correction in the smooth copy of the ITS. The marked-up pages will not be included in this RAI response. ATTACHED PAGES: 1 1 1 None 1 l l l l l l l ? t
1 ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: WC 5.0-003 APPLICABILITY: WC REQUEST: Amendment No.115 dated March 30,1998 revised CTS Sections 6.3 and 6.12 to reflect the merger for the positions of Superintendent Radiation Protection and Superintendent Chemistry into one new position, Manager Chemistry / Radiation Protection. The CTS and ITS have been marked up to reflect incorporation of this amendment. Wolf Creek submitted a CTS license amendment request on July 3,1997 (letter number ET 97-0065) which position title changes of the Shift Supervisor to Shift Manager and Supervising Operator to Control Room Supervisor. In discussion with the Wolf Creek NRC Project Manager on April 10,1998, this amendment request is to be approved in conjunction with the approval of the conversion license amendment request. The CTS and ITS have been marked up to reflect ! the incorporation of this license amendment request. Note that TSTF-258 would eliminate some of the changes requested in the CTS license amendment request. ATTACHED PAGES: Encl. 2 6-1,6-2,6-5,6-6,6-7, 6-23 Encl. 5A 5.0-1, 5.0-5, 5.0-6, new Section 5.7.2 l l
l I ADMINISTRATIVE CONTROLS ! 1 _ 6.1 RESPONSIBILITY gg 4g y 6,.g 6.1.1 The Plant Mana;er shall be responsible for overall
~
l Unit operation and shal delegate in writing the succession to this responsibility during his absence. b N~ l
)
J 6.1.2 Thebiferfisw@ Tea @dh under the ShiftGu69P(p03 shall be respon- 01-01-A i sible _ _ _ _ .for_ the control room command func, tion. ^ -. z;:cc: d:n '::: t: tN:
- m. .m. m m - _. _ _ ru:_ m __ u ._ r_. _ _ _ m _ .. m 2_ _ . .a MNSEEh b h b-E:I JDuring any absence of the 80 from the]
(contrW room while Mie unit is in MODE 1,2,3, or 4, an individual with an active 0102-A
- Senior Reactor Operator (SRO) license shall be designated to assume the -
- . control room command function. During any absence of the SO from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the
_)
- control room command function. I 6.2 ORGANIZATION l I
6.2.1 Onsite and Operatina Corporation Oraanization Onsite and operating corporation organizations shall be established for unit operation end corporate management, respectively. The onsite and operating cu,,,Oisuen orgenizations shall include the positions for the activities affecting the safety of the nuclear power plant.
- a. Lines of authonty, responsibility, and communicatum shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptbns of departmental responsibilities and relationships, and job descriptions for key personnel positions, or equivalent forms of documentation. These requirements shall be documented in the Updated Safety Analysis Report.
- b. The Plant Manager shall be responsible for overall unit safe operation and shall have control of those onsite activities necessary for safe operation and maintenance of the plant.
- c. The President and Chief Executive Officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety,
- d. The individuals who train the operating staff and those who carry out the health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. - ._
6.2.2 Unit Staff au
~ 01-064f, Qhe unit staff organization shall include the following:]
, c~
- a. E:9 :- df; 2*. 9M ' z.- ;:::f O' -' ' _ :'. 'h: - ' 'cr :'?
_m._ :_ .m.- 0107-M _ E211^ nuclear station operator shall be
. [ assigned when fuel is in the reactor and an additional nuclear station operator --
shall be assigned when the unit is in MODES 1,2,3 or 4. WOLF CREEK UNIT 1 6-1 Amendment No. 4.34.46768.100 Mark-up of CTS 6.0 $/15/97
ADMINISTRATIVE CONTROLS Unit Staff (Continued)
- b. A4!::d ::: !!:: ::d 0;:::ter :he!! be !- % c.. ;' reer "; hen 'u ! !: ! 'h: :::dcr. ~ 01 45-A7
% :f 52 , ;'!!: the L'n"!:! "005 *,2,2 er ', et !: d One !! ented Sen!: 0; : 'er . .G <w.
05:!!5:! *:::n' !:::r;
- c. An individual from the Health Physics Groups *, qualified in radiation protection procedures, shall be on site when fuel is in the reactor;
- d. .^LL C^"E .^.LTE ^7?ONS the!' 5: Ober :d :nd dire sj rup: ;!::d bj - 0143-A'"
3
- "'.: :!!::n::d S;n'r O;; :^ r ::!!:: ::d F:n!:r 0; rd - L! ":d 2wMM t F u ! " :nd!!n; ; t : h:: n Other -^ncur:n' m:; n:!b!':t!:: dudng
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- -t:!'"n ": .: dudng : 5 : ;- :r;:nr;; :nd
- f. Administrative procedures shall be developed and impiamented to limit the working hours of Unit Staff who perform safety related functions; e.g., Senior Operators, Operators, Health Physicists, Auxiliary operators, j and key maintenance personnel. __
IThe unt -overti wo rforminglafety / t-.co-A by UgStaff~' fembe l fu ibeI i NRC Bdlicy StaNimept Q 5 2-1 (Gene - e Mn --[ lhl SE e.T 2. - 2m , 2 i l
- g. The Superintendent Operations or Manager Operations shall hold a senior reactor operator license.
*May be less than tN minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.
WOLF CREEK UNIT 1 6-2 Amendment No. 24,-64,81 Mark-up ofCTS 6.0 S/ ISM 7
. _. . _. . . . - ~ _ . .. _.- ~... _._. _ - .- _ . - . - .~.
T-AEkE.4i-G-4 01-06-LG l
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l loc, G.o-ou } ! ST^ CP Sh* T:2.. '- ' ^f;!:: . . E - r ,, - '"---' 2P " "") l C'.;mi:^ / F:xxx'. i ! The Shift Crew Composition may be one less than the minimum requirements of [10CFR50.64(mH2HO and 6.2.27:t'- S.21 for a period of time not to exceed l 2 hours m order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composebon to within the minimum requirements. : S.2
- 9: ;z;'r':n '::: n ' ;r- =/
i f.*. r;.. ;:f: . '- h unT. .nnd r;x CJ . -'.x;; t: t: - nn ';.; r.* _ ! r:;c u n i::n; !:^: :: ':- ' gyge s permevQ During any absence of t hm the controlroom while the {tXS.0-603} unit is in MODE 1,2,3, or 4, an individual with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the@spenaswvn Qa rat 0cfrom the control room while the Unit is [ in MODE 5 or 6, an individual with a 'vahd Operator license shall be designated i to assume the control room comrnand function r i - f u_
% Yg;g
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i t i s WOLF CREEK UNIT 1 6-5 Amendment No. 54 Mark-up ofCTS 6.0 SMS/97
l 1 ADMINISTRATIVE CONTROLS l 9 2 2 '"DFPF"OE"T F ^".F"G!NEEo'"" GP.OUo cE" 01-04-LG l er ,eu. _rv. i.n_ u . ! i S 2.2.4 The .'SEC th Rndien te ::m:nc p!:n! cperating 2 rceterbtb , l u o. r. :,. ,_ ._. _ . ... .. : a . .. . m .a... .:.
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"'!c .cy ind!=te ctent f mprev ng p!:^t =fe'y. 'rhe !SEC 2:!! .20 l
deta!!cd r^^^mmendatient 'cr reviced preceduret,0 ;uipmen* med!=tient, mafr
!cn ne 20!!":t!::, Oper t:ent :2"tb: cr c!her cen: Of:mprev:ng p! nt
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. - m ev 6:2.2.2 The !SEC 2 !! be repentt!: for mainte!r:n; cur;;!!!:n= c' p!:n' .~ a . .:--
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6.2.4 SHIFTTECHNICAL ADVISOR As w, Aw, W ,4 . T ch or all Provide technical support to theliMrbudeadODin the areas of thermal hydraulics, reactor engineering and plant ana[ sis with reaard to the safe operation of the Unit.
--(Lent operhtiew.s sWtt cntC k N'I )l 6 3 UNIT ST FF QUALIFICATIONS -
6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI /ANS 3.1-1978 with the following exceptions:
- a. Licensed Operators and Senior Operators shall meet or exceed the qualifications of ANSI /ANS 3.1-1981 as endorsed by Regulatory Guide 1.8. Revision 2.
*"c :::p m!b!c for :!gn c"'und:en . 34-LG is
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j Q Et-l - M :m > ev- uc5.O-003} WOLF CREEK UNIT 1 66 Amendment No. t .* ? 4 d, S4 Mark-up of CTS 6.0 S!!S/97
l ADMINISTRATIVE CONTROLS __ - -
' 6.3 UNIT STAFF QUALIFICATION (Continued) hW8' "*"M l)
- b. The position ofDeptionnipoenmadcation Protection shall meet or C 5600h exceed the qualificatons of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager.
,.u. . . u. e. o. r_ _ _ m _ _ _. . . _ ._. u._u.. . _ . . - _. . _ __ . _ _ _ _ _ _2 m. . , . _ _ . .:. _ _ _ _ . _ _, A M. e !.f u.t e. # ggg{g ?
, 4
.. 4. n 4
- d. The position of Manager Operations shall hold or have previously held a senior reactor operator license for a similar unit (PWR).
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e ._.y I i r i I WOLF CREEK 6-7 Amendment No. 20,25,i5.51,5",70, S t ,100 Mark-up of CTS 6.0 S/l$/97 i
- - - . . . _ - --_ . . - . 4 _
.e ]
l ADMINISTRATIVE CONTROLS I S ' o ^ O!.^ v!OM o OTSOT!OM noeco^"
. 0310-LG i "re dur: f0r p::::nn:: :t ten prte:P n th " 50 pr; ed cent!: ten' i
"' 'h: m;r..;nn:: :! 10 CFo P -t 20 2nd 05:!! bc p;m :d, .cintrin:d end
- Ch:.;d te f:::!! :;; ::-.;I ;&:n;; r:n :' rf S:n cr; :u :. I 6.12 HIGH RADIATION AREA 03-11 s )
6.12.1 Pursuant te P r;rph 20.202( )(5) of 10 CFR Part 20. in lieu of the "Z...J fit" 0 "1-:^^ d--d" :^^td bl 2:2;;-~~". 20.202( }l2[ f_r_eements of 10 CFR 20.1601] each high radiation area, as defined in i 10 CFR Part 20, in which the intensity of radiation istreater than 100 mrom/hr} (Q6,1-1I but equal to or less than 1000 mR/h st 15 := (19 h liWcmQoW*rr .: rf: 1
!! n Cur: Or frr crj tuM :: r".!:h 'h: ef ' en ;;n:.._ :: shall be barri-
._caded and conspicuously posted as a high radiation area and entrance thereto ,
shall be controlled by requiring issuance of a Radiation Work Permit (RWP). l Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or norsonnel continuously escorted by such individuals may be - exempt from the RWP issuance requirement during the performance of their assigned duties in high radiationareas with exposure rates equal to or less than 1000 mR/hbt 30 cnOErG[ provided they are otherwise following plant
~
l Q S. 2.-l J
. radiation protection prococ ures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which contiriuously indicates the radiation dose rate in the area, or '
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or
- c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform penodic radiation surveillance althe frequency specsfle(btaHE SM 't^^:C^'.O-i ^ML apffysickj % E1 I } -
$3 Il}AI
(= , m . Q
.i n The RWP.
i - g_- . _ n 6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible - to personnel with radiation levels greater tharfor equal tol1000 mR/h at 46 em 414-in4 03:1FA ^ [30 cm t12 in.Bhcr er; ::M::: N:'. '5: met:n ; n:: ter ahM shall be provided with locked doors (or continuously guardedMo prevent +neuthenaed *" 03-2643-3' (IEC1"-Santry. and the keys shallbe mamtained under it.e administrative control of the h ~m1
'supar)paer/gypervisinrQper@on duty and/or health physics supervision.
~ Doors shan remain locKea except during periods of access by personnel under 1 an approved RWP which shall specify the dose rate levels in the immediate work [ cp E. 2.-l } .
areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the KWP, direct or remote (such as
- closed circuit TV cameras) continuous surveillance may be made by personnel l qualified in radiation protection procedures to provide positive exposure l
control over the activities being performed within the area. _ L Shdt Mmadjer/Co koqqm*Nov- -lwc. 5.0 -00.% l 7/OLF OREEK UNIT 1 3-23 Amenomem No. 4-4 81 liarA-up of CTS 6.0 S/15/97 s/-
, - , ,,, - - - - - . , - - , , - . - , , , , , , , - - , , , , . , - - , . - + . . , - , , , , , . , -
Responsibility 5.1 5.0 ADMINISTRATIVE CONTR9.S 5.1 Responsibility 5.1.1 The Plant Superintcadcat Plant. Manager shall be responsible for 1B;PSBig overall unit operation and shall delegate in writing the succession to this responsibility during his absence. The [Plent Supcrintcadcat] cr his dcsigacc shall opprovc, prior to . inpic;catetion, each proposed test, experi;;at or m,r,dification to d6El*W$ Nna@ systcas or cquipcent that affcct_ nacic i g rom
;fetv-c <r sup7ermdyy 5.1.2 The Shift Supervisor 'SS) uper9 M M under1the; Shift
@d8?'PSites
- soMshall be responsible for the control room command " Gnbl Raam I function. During any absence of the SSJur%cvfsi>ig pgferayor)from the a
^"g;p37"a
'hper v d.d control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SR0) license shall be _ designated to assume the control room comand function. During any absence of the SStu#ryKimt6B from the control room while FB+PS) ~ the unit is in MODE 5 or 6, an individual with an active SR0 license ~ ' or Reactor Operator license shall be designated to assume the control . room comand function. l
)
t 1 1 i i i l 1 i 1 4 l l WCGS-Mark-up ofNUREG-1431-1TS S.O 5.0 1 S/15/97
- . . .- . . . . _ - . ~- . _ . . .
Organization 5.2 5.2 Organization 14 s.2.-tj gf. M erdhi d d N N S M sha jlprovide advisory technical ^; " * : support to the.(fbut sapp >(EdrA5D1n the areas of thermal wc s.o-m] hydraulics. reactor engineering, and plant analysis with regard to the safe operation of the unit. ............~.m . . . . . FAposi_ tion ~s. hall .be_ manned;jniHODES 122;, lor;4j :unless #A theRor. the.jndi~vidual.with a SeniorJ0perator License meets GNp mMcD the qualifications specified by the Comission Policy Statement on Engineering Expertise on Shift. I j l l I WCGS-Mark-up ofNUREG-1431-ITS S.O 5.0 5 S/1S/97
- -- =- . . . _ _ .. . _ - - . - . . . _-. - - . _ _-.-
Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications Revicx r'; i;ctc. tiinis.um qualific; tion; for n;..Mr; cf the unit -taff ; hall bc
;pccificd by a- of an everall quali'ication -t;t;;;nt rcfcrcacing on "JCI St;nd;rd acccpt;bic t; the tRC ;teff or by pccifying individual p;;ition qualific; tier.;.
Ocacrolly, thc fir;t =thod i prcierabic, henscr. the ;ccend =thod i; edoptabic to th; c unit -teff rcquiring ;paial qualification ;tatc;. cat- bacu c of unique
,rg;nizationel ;tructur ;. -
5.3.1 Each member of the unit staff shall meet or exceed the minimum g~5;34' ~ ' qualifications of ANSI /ANS 3.1 1978 with the following; exceptions:
~~
[ Regulatory Guid; iue, cm ._; _ a _____m.u,_ 1.0, Rcvi; ion 2.1^07. cr ser; rcer.t revi; ion;. or jg;psjg i_ mu_ in, . .,,, ru_ _ .,, __. __..__a t.. rwisa .4 wus ruu s u eswwbybuweb hv &ssw r is w abug aJ. sasb abu4 a s rv b hv T UI ww WJ [Rc ul; tory Ouide 1.0] ; hell x ct er := a d th; minimum Pfication; cf [Rcgul; tion;. Regulatory Guidc;, or "J5: Stand rd-ee n._m.<,. emum.m em si.m, m, emu. 2 5?3 @ l Liceris~e d10perators:and_SenioE0peraldrsZiha]1Xueet; Eor
$R33*i$
exceedithe qualifications of ANSI /ANSJjl;1981_ as; endorsed byjRegulatory Gui.de 1.'8,;Revisfon T and'10~CFR_ Part 55.
;-Q1an%E chem' @ C 19c ? 0-23 \ l 5.3.1.2 Thelposi_ti; n . ofM_ a_f..h eQRadi at.1 on : Protecti_on_shal.1 '
me.e.t".o..~r: exceed the qualifi. - - cati..ons~of. ~R.egul.atory Guide 1.8, . September 1975~for a Radiation Protection: Manager. I 5.3.1.2 The position _ of Manager Operations,.shall_ hold.or. have previcasly held a senior reactor operator. license for a similar unit (PWR). 5.E 2 ib- the pr ou. of 10 CFR 5 5.44, o. h c.nmed. Semin. Lxhn. Operatut (5 D) M 2 licasad reachoq e 2.to v- ( Rcd c v e. t b osa vA chvi dels do, A a acute to me.5- 3 k *
.th.4. <qde %ts d TS 5.3. i, .pu'fdo" h fwn c t tous '
da.scA b4A. b 1c) ccR SD 54(M .
~
1 l i i I 1 l WCGS-Mark-up ofNUREG-1431-ITS S.0 5.0 6 5/15/97
_ ~. .. . _ . _ _ . _ . __ _ _ _ _ _ High Radiation Area 5.7 I ! 5.7 High Area Radiation Area I 5.7.1 Hiah Radiation Areas with Dose Rates Not Exceedina 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: (continued) , 4. A self-reading dosimeter (e.g., pocket ionization chamber l or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the l area; who is responsible for controlling personnel l exposure within the area, or (ii) Be under the surveillance as specified in the RWP or
- equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
- e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.
I 5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation:
- a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked er continuously guarded door or gate that prevents unauthorized entry, and, in addition: gT j Q y g p.sc-ca31
- 1. All such door and gate keys shall be g maintained under the administrathe control of theMMLegoerstforgfacat>on
! pyotet).tetryanager), or his or her designee.
%-Ps
_healkh phyics supervimo9 kj (continued) l l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 5.0-004 APPLICABILITY: WC l REQUEST: Amendment No.106 was issued on June 24,1997. This amendment modified CTS 6.8.5.b to provide an exception to the examination requirements of Regulatory Guide 1.14, Revision 1,
" Reactor Coolant Pump Flywheel Integrity" and delays the inspection of the "D" reactor coolant pump flywheel to the Fall 1997 refueling outage. As identified in Attachment 3 to the conversion application, the amendment request was incorporated into the conversion application.
Subsequent to the submittal of the conversion application, the license amendment request was revised to include a one-time allowance tc extend the performance of "D" reactor coolant pump examination. This licensee identified item reflects in the conversion application the approval of Amendment No.110 to the Wolf Creek license. New DOC 2-21-A was generated to delete the one-time extension for the "D" reactor coolant pump examination. DOC 2-21-A states: i
" Amendment No.106 for Wolf Creek incorporated a footnote to allow the volumetric and surface l
examination of the RCP "D" motor flywheel for the first 10-year ISI interval be delayed for one operating cycle. The examinations are completed during the ninth refueling outage. Since the footnote is a one-time exception and has been satisfied, the footnote is no longer applicable and can be deleted." i ATTACHED PAGES: Encl. 2 6-18c j Encl. 3A 7 Encl. 3B 5 l
l ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 3. A surveillance program to ensure that the quantity of radioactivity contained in following outdoor liquid radwaste tanks that are not surrounded by liners, dikes, l
or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains ! connected to the liquid radwaste system, is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table @4, Column 2 at 02-09-A the nearest potable water supply and the nearest surface - water supply in an UNRESTRICTED AREA, in the event of an . uncontrolled release of the tanks' contents
- a. Reactor Makeup Water Storage Tank. ,
- b. Refueling Water Storage Tank, l
- c. Condensate Storage Tank, and
- d. Outside Temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste. ,
The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.
- b. Reactor Coolant Pumo Fivwheel Inspection Prooram Each reactor coolant pump flywheel shall be inspected per the remis.eadations of Regulatory Position C.4.b of Regulatorv b2M
- Guide 1.14, Revision 1, dated August,19757 in lieu of Position C.4.b(1) and C.4.b(2), conduct a qualified in-place UT examination over L{T.x 5.0-c04. l the volume from the inner bore of the flywheel to the circle of one-half the outer radius or conduct a surface examination (MT and/or PD of exposed surfaces of the removed flywheels once every ten years coinciding with the Insertice inspection schedule as required by ASME Section XI.
- c. Containment Tendon Surveillance Prooram 4
This program provides controls for monitoring tendon performance, including the effectiveness of the tendon corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial plant operation as well as periodic testing thereafter. The Containment Tendon Surveillance Program, and its inspection frequencies and acceptance criteria, shall be in accordance with Wolf Creek Generating Station position on draft Revision 3 of Regulatory Guide 1.35, dated April,1979. The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Containment Tendon Surveillance Program inspection frecuencies. 92-11-M E "new. Technical Specification Bases Control Proaram (see insert 10)' new. Safety Function Determination Proaram (see insert 11) s 02-11-R _ u
,&GW 6.9 REPORTING REQUIREMENTS (The following reports shall be submitted in accordance with 10 CFR 50.4] 03-01-A.
"OU"!ERE" OPT?
S.9.* M ::r '? te th: :pp!:::b!: rep e; re ;u!r =nt: cf"t!: 10,Ced: cf F:d re! ;6t! n:, *he fe!!= n; rep -t: th:!! 5 rubMed t the Pe;!:ne!
^drM!:trter f'h MoCR:;!:ne!0" :u2 : Other !:: n ted.
WOLF CREEK UNIT 1 6-18c Amendment No. 8BrG'6101 i rk-up of CTS 6.0 S/15/97 2.,2g ., a e _; m_ c.__ o . - emu _ ; .. h - - . . A . ., . _ r u., o _;g,,, __ 7,,m y.,"p" -.t.., I' y ' -d.
.:. m ., _ - .
k- ^
#^
'O y q [" "." . .[. . .",^ {. . - .^_' *_ ,. M Y'
, lAlf. S.b.to16 l
2.6 Lcc A %. w_4ss, sec. comessim Comp Wew T* l CHANGE l NUPEER NSHC _ DESCRIPTION f Coz-2CA_ nuestr 3A-7b}--Lu s.o -ooc which reference these programs, and therefore, the lack of an applicability statement in the Programs introduces confusion. _A__ lMEst":r .3A -7a 3 ! o s.2. -I l 03 01 A Revises " Routine Reports" section to be consistent with NUREG 1431. The method _for submitting all reports is revised to be in accordance with 10 CFR 50.4. Since this change merely makes the TS consistent with the regulations, it is considered administrative. 03 02 A The requirement to submit a Startup Report is deleted from the CTS to be consistent with NUREG 1431. This report required no staff approval and was submitted after the fact and is therefore not required to ensure safe plant operation. The approved 10 CFR 50, Appendix B. QA Plan, and USAR startup testing program provides assurance that the affected activities are adequately performed and that. appropriate corrective actions, if required, are taken. 03 03 A The Annual Reports section is revised to be consistent with NUREG 1431 and traveler TSTF 152. Names and formats are revised consistent with NUREG 1431. Also, revises the annual report section to reflect the new 10 CFR Part 20 requirements and associated reconstnded changes noted in I PRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10CFR20 and 50.36a Changes." (From Christopher I. Grimes to Owners Groups Chairs). ) l 03 04 A The requirement to report specific activity limit ' violations is deleted consistent with NUREG 1431. This report is a history of Reactor Coolant System (RCS) l specific activity Limiting Conditions for Operation (LCO) entries. GL 83 43 and revised reporting requirements in the regulations intended that LCO entry reports no longer be required. The reporting requirements in regulations cover situations such as seriously degraded barriers (fuel failure). Therefore, every violation of the RCS specific activity LCO need not be reported. Serious degradation of a fission product barrier, among other more serious events are required to be reported by 10 CFR 50.73. This change is administ,'ative in that it only affects reports and do not affect plant operations. 03 05 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 03 06 A CTS [6.9.1.7], " Annual Radioactive Effluent Release t Report" and CTS [6.14.c] is revised consistent with NUREG-WCGS-Description of Cisanges to CTS 6.0 7 S/15/97
.- _- - - _ - - r - . , . , -,y-
t INSERT 5.0-23 0 5.5-7
- f. Demonstrate at least once per 18 months for each of the ESF systems that following the creation of an artificial Delta P across the combined HEPA filters, the prefilters, and the charcoal absorbers of not less than the value specified below (dirty filter conditions), that the flowrate through these flow paths is with i 10% of the value specified below when tested in accordance with ANSI N510-1980.
ESF. Ventilation System Delta P Flowrate Control Room Filtration System 6.6 in. W.G. 2000 cfm Control Room Pressurization System 3.6 in, W.G. 750 cfm Auxiliary / Fuel Building Emergency Exhaust 4.7 in. W.G. 6500 cfm 9 .r, M h ._a
{ Att: chm:nt 3 to ET 98-0078 l P:ge 1 of 1 l LIST OF COMMITMENTS l l Th3 following table identifies those actions committed to by Wolf Creek Nuclear Operating l Corporation (WCNOC) in this document. Any other statements in this submitta.1 are provided for ! information purposes and are not considered to be commitments. Please direct questions reqarding th:se commitments to Mr. Michael J. Angus, Manager Licensing and Corrective Action at Wolf Creek
- Generating Station, (316) 364-8831, extension 4077.
t COMMITMENT Due Date/ Event l A supplement to Reference 3 (ET 97-0050 - conversion Prior to issuance , license amendment request) will be submitted at a later of SER dr_te. The Withdrawn Specimen Test Results Report wi!! be submitted 9/30/98 to the NRC by the end of September 1998.
~
7 l \
l RCS Loops-MODES 1 and 2 B 3.4.4 i B 3.4 REACTOR C00LAKr SYSTEM (RCS) B 3.4.4 RCS Loops-MODES 1 and 2 BASES ' i BACKGROUND The primary function of the RCS is removal of the heat generated in the ; i fuel due to the fission process, and transfer of this heat, via the ' steam generators (SGs), to the secondary plant.
)
! The secondary functions of the RCS include:
- a. Moderating the neutron energy level to the thermal state, to increase the probability of fission:
- b. Improving the neutton economy by acting as a reflector:
- c. Carrying the soluble neutron poison, boric acid;
- d. Providing a second barrier against fission product release to the environment; and
- e. Removing the heat generated in the fuel due to fission product
~-
decay following a unit shutdown. ndlav. S .+. Gen - d The reactor coolant is circulated through ou loops connected in parallel to the reactor vessel, each contain ng an SG. a reactor coolant pump (RCP), and appropriate flow and temperature ; instrtmentation for both control and protection. The reactor vessel contains the clad fuel. The SGs provide the heat sink to the isolated secondary coolant. The RCPs circulate the coolant through the reactor l vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the reactor coolant ensures mixing of the coolant for proper boration and chemistry control. APPLICABLE Safety analyses contain various assumptions for the design SAFETY ANALYSES bases accident initial conditions including RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LC0 is the reactor coolant l forced flow rate. which is represented by the number of RCS loops in l service. l i (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4-17 5/1/5/97
_~_- - - - . _ - - . - . . . . . ~ . - . . . . .- ._- - . - . . - - - . - - - - . - - RCS Loops-H00ES 1 and 2 i B 3.4.4 BASES (continued) APPLICABLE The plant is designed to oprate with all RCS loops in operation h'h-SAFETY ANALYSES to maintain DIER above the41mitival_ues, during all normal operations 1 (continued) and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel I cladding and the reactor coolant. l RCS Loops-H0 DES 1 and 2 satisfy Criterion 2 of th; !"'O ."clicy ) l St;t a c. .10 CFRl50:36(c)(2)(ii). LCO The purpose of this LC0 is to require an adequate forced flow rate for I core heat removal. Flow is represented by the number of RCPs in operation for removal of heat b the SGs. To meet safety analysis i acceptance criteria for DNB, four ptmps are required at rated power. c m lo s. aen-ei _ __ LNh I An OPERABLE RCS loop consists _of an OPERABLE RCPfja opiu:atTon-proy4 ding fopeird flotfor-heat-teensper_t)and an OPERABLE SG_in accordance wTth_ Steam Generator Tube Surveillance Program /AWReF'ishmE !
;+i o.s cyak.sa. of caa 3 peu meas mie akIa. to pro vid n. tseenat '
# leu.s . _
4 APPLICABILITY In MODES 1 and 2, the reactor i; critic;l ;r.d thu; when criticalJhas the potential to produce maximum THERNAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DtB and core damage. The decay heat production rate is much lower than the full powe.' heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for H0 DES 3, 4, and 5. Operation in other H0 DES is covered by: LC0 3.4.5, "RCS Loops-H0DE 3": LC0 3.4.6, "RCS Loops-H0DE 4": LC0 3.4.7, "RCS Loops-H0DE 5. Loops Filled"; LC0 3.4.8 "RCS Loops-H00E 5, Loops Not Filled": LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (H0DE 6); and LC0 3.9.6, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (HODE 6). L (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 19 S/1/5/97
..-. - - _ . - -. . _ - . . - .. . -.- - - ._ .- - - .~. .-- - _ - - -
RCS Loops-H00E 3 B 3.4.5 1 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.5 RCS Loops-H0DE 3 BASES l BACKGROUNC In H0DE 3, the primary function of the reactor coolant is removal of l decay heat and transfe of this heat, via the steam generator (SG), to l the secondary plant fluid. The secondary function of the reactor ' coolant is to act as a carrier for soluble neutron ooison. boric acid. radhvie Q 3.4.(ses-s i The reactor coolant is circulated through ou RCS loops, connected in parallel to the reactor vessel, each containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure, level, and l temperature instrumentation for control, protection, and indication. The reactor vessel contains the clad fuel. The SGs provide the heat sink. The RCPs circulate the water through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. In H0DE 3. RCPs are used to provide forced circulation for heat removal l during heatup and cooldown. The MODE 3 decay heat removal requirements are low enough that a single RCS loop with one R running is sufficient to remove core decay heat. However, two CS loops are required to be OPERABLE to ensure redundant capa ty for decay heac removal. redtine p,g; APPLICABLE Whenever the reactor trip breakers (RTBs) are in tim closed SAFETY ANALYSES position and the control rod drive mechanisms (CRDMs) are energized, an inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a malfunction of the rod control system. In addition, the possibility of a power excursion due to the ejection of an inserted control rod is possible with the breakers closed or open. Such a transient could be caused by the mechanical failure of a CRDH. Therefore, in MODE 3 with R"; ir, tt.: ci;;;d pc;iticr. ;r.d the Red Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be OPERABLE and in operation to ensure that the acciden analyses limits are met. For those conditions when m\hht NLUn\ \ l t l I (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 21 S/1/S/97
1 RCS Loops HODE 3 B 3.4.5 BASES APPLICABLE the Rod Control System is not capable of rod withdrawal, two SAFELY ANALYSES RCS loops are required to be OPERABLE, but only one PCS loop is ; (continued) required to be in operation to be consistent With H00E 3 accident ' analyses. e Th~~lopierationloCone RCP]n1MII4[andTprovidesi][Mjyowito ensurelsiginggpreyentistratificationgandmoducemiztgitsgety ' changesidutipg_RCS;boronfconcentrationireducti_ons Gt!sMgtor 4 corlantiloopMoperation1 1 nleither!MODESJ3ZMMuronWho ' utstMiterminatediandidilution;sourcesJsoljatedKMbogfld]Rtqon ; aalysts;inithese;M00ESitakelcrgdit2fotthe i Whavingjat]3e_ast:one' yea _ctor[cooinndic,JigggyoDgei orMnMqLorgtggtg - ~ . l Failure to provide decay heat removal may result in challenges to a l fission product barrier. The RCS loops are part of the primary success ! path that functions or actuates to prevent or mitigate a Design Basis i Accident or transient that either assumes the failure of, or presents a ! challenge to, the integrity of a fission product barrier. RCS Loops-MODE 3 satisfy Criterion 3 of tra =0 Policy Stet;;42. -10 CFA,50;36(c)12)fii).. rwt i eru. (cp.2. 4. Cre n - i j LCO The purpose of thi LCO is to require that a east two RCS loops be OPERABLE. In 3 with the RE : in tra " Control Syst apable of rod withdrawal,RCS h;;dloops p;; mon must be ;nd in Rod operation. RCS loops are required to bG in operation in H00E 3 with the RE: ci;;;d ;nd Rod Control System capable of rod withdrawal due to the postulation of a power excursion because of an inadvertent control rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met for all of the postulated accidents. L'ith th RE: in tre sp;n p;;ition, or th CPO; de crar;;i;;d, When the Rod Control System is not capable of rod withdrawal ther fsrc only one RCS loop in operation is necessary to ensure removal of decay heat from the core and homogenous boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to ensure that safety n ly;;; limit; er; retredundancy for heat' removal is maintained. (continued) ) WCGS-Mark-up ofNUREG-1431 - Bcses 3.4 B 3.4 22 S/VS/97
RCS Loops MODE 3 B 3.4.5 BASES ACTIONS C.1 and C.2 (continued) If the required RCS loop is not in operation,'and the n, . . . . s .m end Rod Control System is capable of rod withdrawal, the Required Action is either to restore the required RCS loop to operation or to pTace ~the Rod:Contro11 System;1_nla' condition jncapablegrodyithdrawal (e;g. gby;delenergizingl& crergi n all CRGMs. by opening the RTBs or de energizing the motor generator (MG) sets). When the R5 s ;r; in th; ci;;;d p;sition and Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation, the~RodiControl System must:belrendered~ incapable 1of; rod withdrawal. th; R E ; e st k sp; red. The Completion Times of 1 hour to restore the required RCS loop to operation or defeat the;R_od1 Con. trol System de cacrgin 11 C"J"t: is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period. D.1. D.2. and D.3 rdW [c .s.w.sen-i l If our RCS loops are inoperable or no RCS loop is in operation, except as during conditions permitted by the ilote in the LC0 section,Jace thelRod;C6ntroTSyste(inTa:lconditionLincapable oCrod;Vrithdrawal (esgmbWde-energizing all CRDHs, =;t h de crergind by opening the RTBs or de energizing the MG sets). All operations involving a reduction of RCS boron concentration must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated. Addition;ofLborated water with a; concentration greater. than:orsequalito:the ministmirequiredlRWST concentrationbutlessthanthe_ actual.RCSboronconcentration:shall nottbe1 considered .a reduction:in; boron 1 concentration.. (RefE2): Boron dilution requires forced circulation for proper mixing, and defeating the Rod: Control l System ep;ning the RE: 07 de crarsicing th; Z ;;t: removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued unti' one loop is restored to OPERABLE status and operation. SURVEILLANCE REQUIREMENTS SR 3.4.5.1 , g,g ,g. q =
- This SR requires verification every 12 hour that the required loops are in operation. Verification may' include flow rate, temperature, endor pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient l
(continued] WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4-25 S/1/587
RCS Loops-MODE 4 B 3.4.6 1 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.6 RCS Loops-MODE 4 BASES BACKGROUND In M)DE 4, the primary function of the reactor coolant is the remofal of decay heat and the transfer of this heat to either the stean generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchar.gers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid-l redhnad - lQ.s.w.asa-i 1 l The reactor coolant is circulated through four RCS loops connected in parallel to the reactor vesse , each loop containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure. level, and temperature instrumentation for control, protection, and indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and to prevent boric acid stratification. In MODE 4, either RCPs or RHR loops can be used to provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one R}R loop for decay heat removal and transport. The flow provided by one RCP loop or RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that two paths be available to provide redundancy for decay heat removal. APPLICABLE In MODE 4. RCS circulation is considered in the SAFETY ANALYSES determination of the time available for mitigation of the accidental boron dilution event. The ",00 end "J:", ic~p; provid; thi; circul;ti;n. The operation _of'one RCP in MODES 3,' 4,1and;5. provides adequate flgtoiensure mixing,' preventistratification,1and(produce gradualtreactivity changes during RCS boron 3v&.cgtration reductionsCWith no; reactor coolanthloop?inioperationiin. either MODES ~3,14, orc 5'iboron~dilutionstmust be~ terminated.;and dilution sources. isolated. ;The; boron dilution analysis.inithe.s; MODES. take creditiforgthe mixing vol.ume associated _with havingLatieastLone reactoricoolant ~1_oop_.in operation.;(Ref. f1).. RCS Loops-MODE 4 h;;; b;a idcatified in th; N",0 "clicy St;t;. at ;; i;;;pertent ;;ntrikter; to ri;k reduction. satisfies criterion 4.of'10 CFR 50.36(c)(2)(11)l. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 27 S/1/S/97
1 l RCS Loops-H0DE 4 l i B 3.4.6 BASES LCO The purpose of this LCO is to require that at least two loops be OPERABLE in MCDE 4 and that one of these loops be in operation. l The LC0 allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. An additional loop is
- required to be OPERABLE to provide redundancy for heat removal.
Note 1 permits all RCPs or RHR ptamps to be; removed;frosioperation
& car;i;ed for s 1 hour per 8 hour period. The purpose of the Note is to permit tests that are retJiliWitoJNNiBMolinifdWith60t fTEoCpump;~ noise i;;igr.;d t; v;inct; v;rica; ecciir.t =;1yx; v;12;;. Ora ;f tM tat; grfered iring tM ;;t;rtup tatin; progra i; tM v;;1idtion of red drop tin;; irin; ;;1d i s..ditic .;. Mth with sd itMat flow. ";; n; flew tat my k g riera d in "00C 3. ?. er 5 er.d requir;; t Mt t M pu g M
;teppd for ; :;Mit pri;! cf ti.c. TM Mt; pr it;; tM i car;i;in; cf tM p;;;;p in order *; grfe ;;; thi;; tat ud v;;liit; tM n;md m;;1y;i; vela;;;. If ;Mr.;;;; cre ;;ei t; tM i "00 tMt euld can ; ;M..;,. t; tM fl;; ;Mrsteri;; tin ;f tM
.::, CM 1..,ut = M = ;;; M :=?i1 2: by ur. ;; ting CM :=t l ege4*- The 1 hour time period is adequate to perform the riCrejeisdrftesting, and operating experience has shown that boron stratification is not a problem during this short period with no i forced flow. '
Utilization of Note 1 is permitted previded the following conditions are met along with any other conditions imposed by initi;l ;t;rtup test procedures: l
- a. No operations are permitted that would dilute the RCS boron concentration, therefore maintaining the margin to
! criticality. Boron reduction is prohibited because a l- uniform concentration distribution throughout the RCS l cannot be ensured when in natural circulation: and
- b. Core outlet temperature is maintained at least 10*F below saturation temperature, so that no vapor babble may form and possibly cause a natural circulation flow obstruction.
l nbk a s.s. aes-Q l Note 2 r res tnat tne Secondary side water temperature of each SG be s 5 F above each of the RCS cold leg temperatures before the star of an RCP with any RCS cold leg temperature s u.w.i g 368'F. This restraint is to prevent a low temperature i overpressure event due to a thermal transient when an RCP i started. l i i 2 (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 2B S/1/S/97
1 I RCS Loops-H00E 5. Loops Filled i B 3.4.7 ;
- BASES stratification is not likely during this short period with no LCO (continued) forced flow.
4 1 Utilization of Note 1 is permitted provided the following conditions are met, along with any other conditions imposed by iritial :t;rtu;; test procedures: i a. No operations are permitted that would dilute the RCS 2 ' boron concentration, therefore maintaining the margin to criticality. Boron reduction is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation: and , g_ b. Core outlet temperature is maintained at least 10*F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours, provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such i testing is safe and possible. l rmE .tas.w. gen-g} l Note 3 r ires that the secondary side water temperature of each SG be s F above each of the RCS cold leg temperatures before the star of a reactor coolant pump (RCP) with any RCS cold leg l temperature s 368'F. This restriction is to prevent a low temperature ov pressure event due to a thermal transient when an RCP is started. w g4 3, u ,n ,y Note 4 provides for an orderly transition from N0DE 5 to H00E 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to NODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops. RHR pumps are OPERABLE if they are tapable of being powered and are able to provide forced flow if required. An OPERABLE SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE in accordance with the Steam Generator Tube Surveillance Program. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 34 S/1/S/97
l i RCS Loops-HODE 5. Loops Not Filled
- B 3.4.8 BASES LCO The purpose of this LCO is to require that at least two RHR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Hea'
- cannot be removed via
! the RHR System unless forced flow is used. A minimum of one i ' running RHR pump meets the LC0 requirement for one loop in operation. An additional RHR loop is required to be OPERABLE to l meet single failure considerations. l l Note 1 permits all RHR pumps to be dc caergind. removed from i operation; for s 11hourJ ;;;inutc; aten ; witching fre.;. era leap t; enetter. The circtestances for stopping both RHR ptags are to be hh / N limited to situations when the outage time is shortfana core [ outlet temperature is maintained @l0*F below saturation I opec Qtr@en; wtan "J:" fer;;d flow i; ;tepped.The Net; pre ! g,y ThefNote2 requires reat:torWesselTWater]everbelabove~th6 pesse17 flange torensure { cp .s. w.een.ij the operating;RIKpump willinot be intentionally deenergized during~mid loop; operations. Note 2 allows one RHR loop to be inoperable for a period of s 2 hours, provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and pc asible. An OPERABLE RHR loop is comprised of an CoERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. f Q3.4.%21 In H00E 5 with loops not filled Mgpjar16rtfcaduPWh APPLICABILIT( this LC0 requires core heat removar and coolant circulation by the RHR System. One_RHR~ loop provides sufficient capability for this purpose. However, one additional RHR ~1oop;is required to be OPERABLE to meet single failure considerations. Operation in other MODES is covered by: LC0 3.4.4, "RCS Loops-H00ES 1 and 2": LC0 3.4.5, "RCS Loops-H00E 3": LC0 3.4.6. "RCS Loops - H00E 4": LCO 3.4.7. "RCS Loops-Hv0E 5 Loops Filled": l LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (H00E 6); and j tCO 3.9.6. " Residual Heat Removal (RHR) and Coolant i Circulation-Low Water Level" (HODE 6). i ( (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 38 S/1/5/97
l RCS Loops-MODE 5, Loops Not Filled l l B 3.4.8 BASES APPLICABILITY Th~eWit. ability!is~ modified ~by'a Note;st~ating?that[entryMnt.o 3 continued) HIEE[5; Loops. Notlf_filedifrom NODE;5l-Loop _sTf]ledlisfnot~ permitted wh1%the[LC051s notimetJ5This: Note specifiesfanfeiceptionlto LC013~.0l4"anrwouldjpreventIdraining the:RCSE.which.lwould eTimincteithFpossibility of SG1 heat remov;alEWhileithe}RFR function was degraded. i ACTIONS 6.l If only one RHR loop is OPERABLE and in operation, redundancy for RHR is lost. Action must be initiated to restore a second loop to OPERABLE status. The immediate Completion Time reflects the l importance of maintaining the availability of two paths for heat l removal. ! B.1 and B.2 If no required RHR loops are OPERABLE or in operation, except during conditions permitted by Note 1, all operations involving a reduction of RCS boron concentration must be suspended and action must be initiated innediately to restore an Rm loop to OPERABLE status and operation. Addition [of@ratidyiMj_fth'a , cdrice ntrat16hTgrWaterJthan orleiquar,tojttiefninlattiireiguiredjmlST ) concentration ~but71ess'than the actuarRCS boroni5 concentration
' ~
j shi1Fnolt'be considered a reduct1onjiribdr6n~6onMntratYoriC(Ref. j 2R Borondilutionrequiresforcedcirculationframfat]least"one RCP for proper mixing so that~ inadvertent .......... ......... .... t M n rgir. to criticality canlbe; prevent.ed m a t net k redwed in this t g Of Oper; tion. The inmediate Completion Time reflects the importance of maintaining operation for neat removal. The action to restore must continue until one loop is restored to OPERABLE status and operation. SURVEILLANCE SR 3.4.8.1 REQUIREMENTS
, {p3,g,csew_, j This SR requires verification every 2 hours that one loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.
(continued) l l WCGS Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 39 S/U5/97 i i
l Pressurizer B 3.4.9 i BASES ! BACKGROUND a loss of single phase natural circulation and decreased I (continued) capability to remove core decay heat. Twogroup.sjof~ backup l ' presser'izerih' eat _erslare'normally poweredivialthe]ClaissJ1ET4'.'1_6kV bu.s.es.TThe heater 1oads ~will?be shed after (safety: injection'or bus 7unoervoltage(signal;and manually ~ sequenced back'onto~ the Cla~ss1Ei4.16kVl buses. 1 APPLICABLE In MODES 1, 2, and 3, the LC0 requirement for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. Safety analyses performed for lower MODES are not limiting. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present. Safety analyses presented in the UFSAR (Ref.1) do not take credit for pressurizer heater operation: however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. The maximum pressurizer water level limitdichlensures;that] i s M bebbleiyists]n~the#eisurizerEsatisfies Criterion 2 of tt.: OC ,"clicy St;t, ..b101CFR350I36)(c1(2)Jii). Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG 0737 (Ref. 2), is the reason for providing an LCO. rdhw LC0 The LC0 requirement for the pressurizer to be OPERABLE with @4 water volume s 1940 1657 cubic feet, which is equivalent to 9 . ensures that a steam bubble exists. Limiting the LC0 maximum operating water level preserves the steam space for pressure control. The LCO has been established to ensure the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions. The LCO requires two groups of backup pressurizer-heaters to be OPERABLE prearinr baeters, each with a capacity a-405- 150 kW, capable of being powered from either the offsite power source or the emergency power supply. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer (continued) WCGS-Mark-up ofNUREG-1431. Bases 3.4 B 3.4 42 S/1/5/97
. - . . . .. . . = .
l Pressurizer Safety Valves B 3.4.10 ! BASES (continued) i BACKGROUND The consequences of exceeding the American Society of (continued) Mechanical Engineers (ASME) pressure limit (Ref.1) could include ; damage to RCS components. increased leakage, or a requirement to ' perform additional stress analyses prior to resumption of reactor operation. l APPLICABLE All accident and safety analyses in the UFSAR (Ref. 2) that SAFETY ANALYSES require safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure. The overpressure protection analysis (Ref. 3) is also based on operation of{hrJee safety valves. Accidents that could result in overpressurizaTion if not properly terminated include:
- uh\\ne. \a s.s.oen-\ \
- a. Uncontrolled rod withdrawal from full power:
u.nde.1c+e-
, _ , W3.4.tD-O ,
- b. " Loss of rcactor ccalant ficw (eedwater line brea -
)
- c. Loss of external electrical load- _
g4.te _: } ;
- d. Loss of normal feedwater: /
["* "b
- e. Loss of all~ non emergency AC' power to station auxiliaries: i and l
- f. Locked rotor.
. 4 e ailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in cycnts c. d.
and c (abovc) the above events to limit the pressure increase. Compliance with this LC0 is consistent with the design bases and ! accident analyses assumptions. Pressurizer safety valves satisfy Criterion 3 of thc NRC Policy Stct: ent. 10 CFR 50.36(c)(2)(ii).
%d clavrx c,*A a.nemMy m)ech@ los.4.io- ) l LC0 The three pressurizer safety valves are set to open at the RCS 3 design pressure (2500 psic 2485 psig). and within the ASME specified tolerance, to avoid exceeding the maximum design l pressure SL to maintain accident analyses assumptions. and to comply with ASME requirements. The upper and lower pressure ,
tolerance limits are based on the i 1% tolerance requirements ! (Ref.1) for lifting pressures above 1000 psig. (continued) i WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4-47 S/1/5/97
i Pressurizer Safety Valves B 3.4.10 BASES (continued) l l LC0 The limit protected by this Specification is the reactor ) (continued) coolant pressure boundary (RCPB) SL of 110% of design pressure. ! Inoperability of one or more valves could result in exceeding the SL if a transient were to occt'r. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS l ' components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation. redun lo s4.cnn-a i APPLICABILITY In MODES 1, 2, and 3. :nd pertion; cf f "^^C 4 ;ke; th; LT^"
; ;;;ing tu v;r;tur; OPERABILITY of hree alves is required i because the combined capacity is requ red to keep reactor coolant '
pressure below 110% of its design value during certain accidents. MODE 3 ;nd pertion; cf "00: 4 er; is conservatively included. l although the listed accidents may not require the safety valves a for protection. The LCO is not applicable in H0DE 4. J.;n all ",00 ;;id 1;;
- t. ver;tur;; ;r; . 320"I er in "0DE 5. or.H0DE~6fwithithe i reactor vessel; head'on because LTOP is previded iniservice. l Overpressure protection is not required in MODE 6 with 3 '
reactor vessel head detenefened NuiWt@T1erftMEki.63qu=w inches). The Note allows entry into MODES 3 end-4 with the lift settings
; outside the LC0 limits. This method permits the;inplace testing and examination of the safety valves at high pressure and 1 temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design j 4 condition. Only alve at a time will be removed from service <
, for testing. The ur exception is based on 18 hour outage time for each of t r alves. The 18 hour period is derived from operating experi
. hat hot testing can be performed in this timeframe.
] g gl {Q 3.4.6ew t ACfl0NS M With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes
- reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve coincident with an RCS overpressure event could challenge the integrity of the pressure boundary.
(continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 48 S/1/S87
l Pressurizer Safety Valves B 3.4.10 l BASES (continued) f l l ACTIONS B,1 and B.2 l (continued) l If the Required Action of A.1 cannot be met within the required Completion Time or if two or more pressurizer safety valves are inoperable, the plant must be brought to a H00E in which the requirement does not apply. To achieve this status, the plant must be brought to at least H0DE 3 within 6 hours and to H00E 4 with any RCC ccid ic; tea.peraturcs . 320*F within 12 hours. The ! allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. With any RCS cold leg temperatures at or below
-a.46- 368'F, overpressure protection is provided by the LTOP
% 3.4.6en. i System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by nree pressurizer safety valves, redhm osscren-ll Addition.to;thelRCS_of borated water with.a. concentration _ greater than or1equalTto the minimum; required;RWSTiconcentrationishall.
not'be considered:a positive reactivity. change. 'Cooldown;offthe RC_SloArgtolation ofioper.$jljtyAaguessyrizenatde,;inde&y
@M9ttManegame!*WintorWrutmN not;be consideredypositiveireactivity;changejpmvided.;theiRCS 1@ratedito .the:. COLD:SillTDOWNRxenon} free condition _per specification'3ElijRefT 5)
SURVEILLANCE SR 3.4.10.1 REQUIREMElfTS SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of Section XI of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No cdditional requirements are specified. Thc picssuriccr safety volve sctpcint is 1 [3]t for OPCPASILI"/. howcvcr. the valvcs arc rcsct to i it during the Surveillence to clicw for drift. REFERENCES 1. ASME, Boiler and Pressure Vessel Code, Section III.
- 2. FSAR USAR, Chapter 15.
l 3. WCAP 7769, Rev. 1. June 1972. l (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 49 5/1/5/97
- -.- - - - . =_. .-. - . _ - . _ . - . --._.- -_ - .
Pressurizer PORVs B 3.4.11 1 l l BASES (continued) l i BACKGROUND Pressure-High reactor trip setpoint following a step reduction (continued) of 50% of full load with steam dump.
~
In addition,the PORVs minimize challenges to the pressurizer
.iafety valves and also may be used for low temperature protection (LTOP). See LC0 3.4.12. " Low Temperature Protection (LTOP) System."
APPLICABLE Plant opentors may employ the PORVs to depressurize the RCS in SAFETY ANALYSES response to certain plant transients if normal pressurizer spray is not available. For the Steam Generator Tube Rupture (SGTR) event, the safety analysis assumes that manual operator actions are required to mitigate the event. A loss of offsite power is assumed to accompany the event, and thus, normal pressurizer spray is unavailable to reduce RCS pressure. The PORVs are assumed to be used for RCS depressurization, which is one of the steps performed to equalize the primary and secondary pressures in order to terminate the primary to secondary break flow and the radioactive releases from the affected steam generator. The PORVs are used also,.modeled in safety analyses for events that result in increasing RCS pressure for which departure from nucleate boiling ratio (DNBR) criteria.ipres~surizer:Volumeror hot.1Qsaturation_are examined (RefE2) or; critic;1. By assuming PORV menuet actuation, the primary pressure remains below the high pressurizer pressure trip setpoint.; thus, tThe
~
DNBR calculation is more conservative and~the transient pressurizer water. yolune is max.imized, and;the hot leg saturation temperature is_ reduced for,those~ transients;assimingiPORV operation. As such, this; actuation _is not; required;to mitigate these. events,and:PORV;automaticoperationjsg;thereforef,inotan assumedjsafety ' function.J"; ;ute;;ti; ;;tu; tion of th; ~",": is ret ;s;cid in any of th; design basis accidents during "00CS 1 2 er 3. Ev;nt; that ;;;;; this ;;r" tion includ; ; turbin; trip, and th; less of reraal f;;d.;;ter 0D.(q 3.u,. Gen-t } Pressurizer PORVs satisfy Criterion 3 of the NPs0 P;1 icy St;t; cat. 10 CFR 50.36 (c)(2)(ii). LC0 The LC0 requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR. l , (continued) l WCGS-Mark-up ofNUREG-1431 - Bases 3.4 6 3.4 52 5/1/S/97
Pressurizer PORVs B 3.4.11 BASES ACTIONS B.1. B.2. and B.3 (continued) ** - lo.u Gm-d If oneI PORV l is inoperable and not capable of being manually cycled, it must be either restored or isolated by closing the
. associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour are reasonable, I based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If
, the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this additional time, the plant must be brought to e "00: in #,i P tra LOO de;; ret ;pply atJ1eastMODEL3:withTm < 500*F, as required by Condition D. C.1 and C.2 If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of I hour or place the associated PORV in manual control. The prime importance for the capability to close the block valve _ _ is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of I hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORV remains OPERABLE, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition B, since the PORVs are not may not be capable of mitigating an ;verpr;;;' ar; event when pix;d in nr.al certrel if the inoperable block valve is not full open. If the block valve is restored within the Completion Time of 72 hours, the power will be restored end to the PORV. r;;ter;d t; OPEPX LE ;teta;. If it cannot be restored within this additional time, the plant must be brought tc ; "^0E in 2,ict, tbc LOO ds; r.;t ;pply at'least' MODE 13'with Tm S 500*F, as
- required by Condition D.
(sekT a 3 4-s ) I 4 3A 4l J i (continued) i ? WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 55 S/1/58 7 1
i LTOP System B 3.4.12 BASES APPLICABLE not occur, which either of the LTOP overpressure protection means SAFETY ANALYSIS cannot handle: (continued)
- a. Rendering all htboth safety injection pumps and one centrifugal charging pump incapable of injection;
- b. Deactivating the accumulator discharge isolation valves in their closed positions or;by venting.the;affected '
acctmulator; and . redime. {o.346m-i.J l
- c. Ci:;;11 sing Precluding st t of an RCP if secondary i temperature is more than 50 F above primary temperature in )
any one loop. LCO 3.4.5, RCS; Loops T HODE 3." LC0 3.4.6, ;
"RCS Loops-MODE 4." and LC0 3.4.7, "RCS Loops-MODE 5. ;
Loops Filled," provide this protection. Operation. bel.ow 350*F but; greater than 325'F. with. all_.centrif_ugal chargingjand; safety; injection pumps OPERABLEiis; allowed;foraupAo 4fhoursGDuring. low' pressure, low temperature operation:all angguciiigfgy,,jajpction actuat.intLsimpalslescent;4entainment M!We!L3NWWNENftssentm mmugntanerttaussungte fagurelofithe;ESF.lactuation:Si rcuittyMlltsultfjg3the startinsfofzat:most oneltrainionsafetyanjecttogone centrafogal:chargtnr pump.C andEonelsafetr injectionfpump C For temperaturestabove?325'F.Jan; overpressure event; occurring;asla result ~ofistarting two pumps;can belsuccessfullyjaitigated by operation;of both PORV's..withoutiexceedjngiAppendi.x G limitU Given1the short^ time; duration -that' this condition 'isiallowed and j the; low probability of aEsingle t failure _causingLan. overpressure i event'during this time, theisingle fail.ure ofia PORVfis'{not l asstmed.ZInitiationf of both trains offsafetylinjection'during l this 4; hour'tineiframe'due'to operatorEerrottor;;alsingle ' failureoccurring during testing ofja redundantichannelf are not considered to;be credible' accidents. Although LTOP:is required to be OPERABLE when RCS~ temperature is less:than.368'F operation with all' centrifugal; charging pumps and_both safety; injection pimps OPERABLE 11s acceptable when RCS temperature-is greater than 350*F. 'Should'an: inadvertent-~ safety injection: occur above 350*F, a _ single;PORVlhas sufficient capacityito: relieve the combined: flow rate of'all!pumpsrAbove 350*FGtwo RCPs and all pressurizer safety valves areirequired to be.0PERABLE.. -Operation offan1RCP .elisinates;the; possibility lof a 50*Ff d1fferencelexisting between indicated;andlactualL;RCS temperature;asLaLresultlof heat; transport effects. LConsidering instrument uncertainties only, an indicated RCS temperature'of 350*F is sufficiently high to allow full RCS pressurization in (continued) WCGS-Mark-up ofNUREG-1431 - Bases.J.4 B 3.4 64 S/1/587
LTOP System B 3.4.12 BASES LCO two: independent means to prevent a pumpistart;in accordance wit.h ) ~(continued) SR T.4.~12.2 l l Note ' 2 ' recognizes , the _. Appli cability overl ap between;. LCO* s ;3.4.12 l and15;2_aMstates that two; safe _ty injectjon pumpsiandf.two centrifugal charging pumps _may be: capable of injectingiintoithe R redhhc [QT%en lj (a) In MODE 3 with any RCS cold leg temperature < 368* F and l ECCS pumps OPERABLE pursuant to LCO 3.5.2, "ECCS-Operating", and (b) For; ap .tol4; hour _s ;afterLentering (MODE 4jfrom NODE 3;or _the temperature.:of ~one,or more.RCSicold legs.; decreases'below 325'F, whichever comes first. Notelstates_thatfone;or more safety, injection. pumps 1may becapabliof_injectingj@jhe RCSiin MODES ~5_and'.'6 when.the RCS Wl6D M Purposie;of _ lngitheldecay. heat ~removar functioni No,terstates;that accianulatot ("' W** ""*1 **D h p 3 An-1.\ g Q> _ the;acc0mulator' pressure;13 RCS; pressure.,for"thetex1st:ngitCS; cold]eg t~emperaturQs3110wed A naximum byith[P/Tli.mit ~curvesyprovided initheiPTLREThisytejpermits the accumulator discharge isolation yalve Surveillance.to be performed only;under these pressurelanditemperatureiconditions. The elements of the LC0 that provide low temperature overpressure mitigation through pressure relief are:
;. Tw; RCS relief velve;, os follows; la. Two OPERABLE PORVs: or A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint,
, and motive power is available to the two valves and their control circuits. (D
- W 'At) reAhr.u Ob. fwo OPERABLE RHR suction relief valve I
l (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.4 B 3.4 68 5/1/58 7
- . ~ _-. . _ ~- .- - - . - - . - . . - - - . . . . _ . . . . . . . . . - . - - .-
LTOP System B 3.4.12 BASES
' LC0 "AnRHRsuctionreliefvalveisOPERABLEforLTOPwhenith (continued) RHR suction isolation valves ord its "Jn ;uction velv; are open, its setpoint is at or between 436.5 psig and 463.5 psig. and testing has proven its ability to open at this setpoint.
Bc. One OPERABLE PORV and one OPERABLE RHR suction relief ve.lve; or f j {414 deel } redime-bd. A depressurized RCS and an RCS vent. (w=&dcAM An RCS vent is OPERABLE when open with an area of a 2- W 2'0 square inches. Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient. APPLICABILITY This LC0 is applicable in H0DE 31when tk tgr;tur; cf anylR_CS coldLleg; temp'erature'isV368'E. in H00E 4 Wka ;re "CS ;;1d leg
~~-
tMihtur; ik2d75]"I1n M00E 5 and in H0DE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits in H0 DES;I E27 asi;3 e k v; 275"F. When the reactor vessel head is off,overpressurizationcannotoccurjWit fuel: .f 1 ded, he' reactp _ ess_e1 _ ad. . be aced _ jthe__sseX~orr 101 co ions = .not" .1.ted Ove ssure rotf ion- ma ta. i .ause?- rces r the 11yf . _. yc g .1
~
.rei Iavai,l .e and . !rea or v.e -
1Thea willi - ft 're ev at L lq 3.4.12-+ } LC0 3.4.3 provides the operational P/T limits for all H00ES. LC0 3.4.10. " Pressurizer Safety Valves." requires the OPERABILITY of the pressurizen safety valves that provide overpressure protection during ODES 1. 2. and 3. erd li00: 4 eks; 320*F. Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid. and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time alltes operator action to mitigate the event. (continued) 1 WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 69 S/1/SR7
LTOP System B 3.4.12 ) BASES
~
l ACTIONS C.1. D.1 and D.2. 0.2 (continued) the maximum RCS pressure for the existing temperature allowed by , the P/T limit curves. ' l If isolation is needed and cannot be accomplished in 1 hour, Required Action D.1 and Required Action 0.2 provide two options, either-ene of which must be performed in the next 12 hours. By increasing the RCS temperature to >075 368'F, an accumulator
@ pressure off)3 648 psig cannot exceed the LTOP limits if tie accumulators are fully injected. Depressufizing the accumulaters
{ os.4.em-q ~ below the LTOP limit from the PTLR also gives this protection. The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the al' awed times. L1 In M00Ei3Mthiany1RCSl cold'legjemperatureji3681 Eor.; MODE 4
- 1. ;rs RCS ;;1d 1;; tWrs.tur; i; a [275 "T, with one required RCS relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within a Completion Time of 7 dayc. Two RCS relief valyts/in any combination of the PORVS;g%.q gg (and the RHR suction relief valveVare required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.
The Completion Time considers the facts that only one of the RCS relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. El The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 7). Thus, l with one of the two RCS relief valves inoperable in H00E 5 or in H00E 6 with the head on, the Completion Time to restore two valves to OPERABLE status is 24 hours. l L l p (continued)
. WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 71 S/1/587
LTOP System i B 3.4.12 l BASES l ACTIONS M (continued) The Completion Time represents a reasonable time to investigate and repair several types of relief valve failures without l exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events. L1 l The RCS must be depressurized and a vent must be established within 8 hours when:
- a. Both required RCS relief valves are inoperable: or
- b. A Required Action and associated Completion Time of Condition A, D E, or F is not met; or reatiru.(backahM*^O b sa een-Q
- c. The LTOP Sys is inoperable for any reason other than Condition A. C, D, E, or F.
g D t a .s. S.cs e n - i l The vent must be sized :t 2e98 2.0 square inches to ensure that the flow capacity is greater than that rcquired for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel. The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of adninistrative control requirements. SURVEILLANCE SR 3.4.12.1. SR 3.4.12.2. and SR 3.4.12.3 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of zero safety injection pumps and a maximum of two one charging pumps l are verified to be;4mapable of injecting into the RCS and the L accumulator discharge isolation valves are verified closed and
- l. ledd cut w1?S power removed from the valve operator.
(continued) l WCGS-Mark-up ofNUREG-1431. Bases 3.4 8 3.4 72 S/1/587 l
i LTOP System l B 3.4.12 ' BASES l SURVEILLANCE SR 3.4.12.4 (continued) REQUIREMENTS l available to the operator in the control room that verify the RHR , suction isolation valves remains open. l The ASME Code, Section XI (Ref. 8), test per Inservice Testing Program verifies OPERABILITY by proving proper relief valve i mechanical motion and by measuring and, if required, adjusting the lift setpoint. I SR 3.4.12.5 The RCS vent of 2 0-98 2.0 square inches is proven OPERABLE by verifying its open condition either: l
- a. Once every 12 hours for a valve that ccr.r.ct bc is not locked , sealed, ototherwise secured ~1nstheiopen. position..
- b. Once every 31 days for other vent'pathsle.g..ifor_ay,ent valve, a valve that is locked, sealed, or otherwi.se secured in position). A removed pressurizer safety valve orlopen manway fits this category.
The Any; passive vent path arrangement must only be open when required to be OPERABLE. This Surveillance is required te-be pcrfcrad if the vent is being used to satisfy the pressure relief requirements of the LC0 3.4.12b.~d. SR 3.4.12.6 The PORV block valve must be verified open every 72 hours to provide the flow path for each required PORV to perform its function when actuated. The valve must be remotely verified open in the main control roomJhis Surveillance is only required.to (h performed if the PORV is being used~to meet'this'ntisfics the LCO. La w a w w ps.S.u-Q , The block valve is a remotely controlled, motor operated valve. I The power to the valve operator is not required removed, and the
- manual operator is not required locked in the inactive position.
l Thus, the block valve can be closed in the event the PORV (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 74 5/1/5/97 l l
1 i l LTOP System B 3.4.12 l BASES I l SURVEILLANCE SR 3.4.12.6 (continued) REQUIREMENTS develops excessive leakage or does not close (sticks open) after relieving an overpressure situation. The 72 hour Frequency is considered adequate in view of other administrative controls available to the operator in the control room such as valve position indication. that verify that the PORV block valve remains open. SR 3.4.12.7 Not Used. Coch ic uircd PJlR ruction rclicf volv; shell bc d;;enstr;ted OPERACLE by vcrifying its PJ R suction volvc and PJ:0 suction isciaticr, volva-erc Opcn and by testing it in accordenc; with the
-Ins;rvic Tcsting Progr;;.. (R;fcr to CR 3.4.12.4 for the PJC suction velic Surv;ill;n;; nd for ; d;scripti;n of th; requirc;ents of the Inservice T: sting Prograr..; This Survciliencc is only perfor.;ed if th PJC suction rclicf v;1v; is bcing used to satisfy this LCO.
$leck d ieet m Iwdn3 ht SR 3.4.12.8 guld kr. laded /*uds.hwp Ww0erM Performance of a COT is required within 12 hours after decreasing RCS temperature to s N5 368'F and every 31 days on each required PORV to verify a7d, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the PTLR allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required.
The 12 hour allowance Feeqeeney considers the unlikelihood of a low temperature overpressure event during this time, ank Q A Note has been a ded indicating that this SR isW equired to be met performed 712 hours after decreasing RCS cold leg temperature l to sN5 368'F. Thc COT c;nnot bc pcrferred until in th; LTOP l "00C when thq_POR" lift ;ctpcint can bc reduccd to th; LTOP _ L sctting te mus e per rme witAfn J2' hour's After] lWC,1A-000 ) PH the l g Q i S. (continued) ! WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 75 5/1/Si97 l
RCS Operational LEAKAGE l B 3.4.13 BASES LCO Total primary to secondary LEAKAGE amounting to 1 gpm (continued) through all SGs produces acceptable offsite doses in the EB accident analysies involving: secondary: steam rdischarge to_the_ atmosphere. Violation of this LC0 could exceed the offsite dose limits for th+sese accidents. Primary to secondary LEAKAGE must be included in the total allowable limit for identified LEAKAGE. l
- e. Primaro to Secondarv LEAKAGE throuah Any One SG l rehms toa.S.cren.o The(00 gallons per day limit on one SG is based on the l assump ion that a single crack leaking this amount would i not propagate to a SGlR under the stress conditions of a LOCA or a main steam line rupture. If leakedage is through many cracks, then the cracks are very small, and the above assumption is conservative.
i APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized. In MODES 5 and 6. LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE. LC0 3.4.14. "RCS Pressure Isolation Valve (PIV) Leakage." measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE. ACTIONS AJ Unidentified LEAKAGE, identified LEAKAGE, or primary to secondary i LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut i (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4.B0 S/1/S/97
RCS PIV Leakage B 3.4.14 BASES l ACTIONS withAii althateAfalve y ha de'g dfd the ili of e (continued) reopadcteMfyst pe it afet {Q 3A.14 -2.) A.1 and A.2 The flow path must be isolated by two valves. Required Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the_ PIVs and must be within the RCPB[or ttfe.)nsi l pressurf porfjorr6fAM __ (qs.w. m. } Required Action A.1 requires that the isolation with one valve must be performed within 4 hours. Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts the operation with leaking isolation valves. Required Action A.2 specifies that the double isolation barr_1er@s.4.iq.-2.] of two valves be restored by(closnfg sgpe-6thSP"Vslys-q0altfted for-tsdatTon-ofJrestoring onc lding the PsCS PIV to;within limits. The 72 hour Completion Time after exceeding the limit allows ~~for..the restoration of the leaking PIV toiOP.ERABLE status. This;timeframe considers the time required to complete the Action and the low probability of a second valve failing during this time period. _ ock of te_d miaq bu.x u. hda tboup jQNe"-4 B.1 and B.2 ' If leakage cannot be reduced the system isolated, or the other Required Actions accomplished, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to H00E 3 within 6 hours and H00E 5 within 36 hours. This Action may reduce the leakage and also reduces the potential for a LOCA outside the contcinment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. l i (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 87 S/1/5/97
1 l l RCS PlV Leakage l B 3.4.14 BASES ACTIONS degraded the ability of the interconnected system to perform (continued) its safety function. A.1 and A.2 The flow path must be isolated by two valves. Required Actions A.1 and A.2 are modified by a Note that the valves l used for isolation must meet the same leakage requirements ! as the PIVs and must be within the RCPB [or the high pressure portion of the system). 1 Required Action A.1 requires that the isolation with one l valve must be performed within 4 hours. Four hours provides time to reduce leakage in excess of the allowable limit and I to isolate the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts the operation with leaking isolation valves.
. Required Action A.2. specifies that.the double isolation *
- b'arrier of 'two valves be restored by closin~g some 6th'er*'
valve qualified for isolation or restoring one leaking PIV. The 72 hour Completion Time after exceeding the limit considers the time required to complete the Action and the- -' low probability of a secon~d valve failing during this time period.
* (qs.w.uw-n]
Ih0 72 h0ui C06ipleti0ii Iime ailur caseeding thi lisili ellows f r the iesius.liun vi liiv leakiny PIV tu viERADLL status.
-Thi: ti :freme cunsiuers ine time requireo to complete this ac t4ca and the hw pret,.'uility vi e secono valve tailing '
.durinn thie norind ID9yici;fer 9 te: Iwu upT,ibns are ~
- ;'*euidgd fne 0.9qnjred A,ggjg3 g,{, }h; 3ggggj g gjgg_
(72 Se"r "estorathn) i: ::pr;prieta if isc1;th; of =
- d v h w;;1d p1;;e tie unii in on enen:1yzed c:nditicn.)
B.1 and B.2 If leakage cannot be reduced, [the system isolated,) or the other Required Actions accomplished, the plant must be l brought to a MODE in which the requirement does not apply. l To achieve this status, the plant must be brought to MODE 3 (continueo) iOG :TS 3 3.4-82 Rev 1, 04/07/95
RCS PIV Leakage B 3.4.14 BASES ACTIONS U (continued) The inoperability of _the RHR Sy;ts autocl;;urc suction isolation valve interlock @ndefs tbs-RHR syctenJso+etio3.watveJs Nc.s4. coq \ inc;poble of i;; latins in rctpon:c to a high prc;;urc corditica ord prcicating could allow, inadvertent opening @c ;;hcMt "a* RCS pressures in excess of the RFR systems design pressure. If the RHP%ycta autocic;urc RHR suctionjsolation~yalve interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivatea % valve within 4 hours. This Action accomplishes the purpose _ of the cutccic;urc functi;P interlock, h m oke. m m )a l wc 3.4.. coq SURVEILLANCE SR 3.4.14.1 REQUIREMENTS -_ _
,rfG M M-1I Performance of leakage testing on each RCS PIV@#soR'ipevp]ve used to satisfy Required Action A.1 cri ",cquired Kction K.Z 1s l required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per
~~~ inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested. one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be
, g % Os Acen-i]
Testing is to be performed everthmon(ths, a typical refue cycle if the plant does not go into MODE 5 for at least 7 days. g The d8kFrequency is consistent with 10 CFR 50.55a(g) M 3&otD \ (Ref M as contained in the Inservice Testing Program, is within the frequency allowed by the American So iety of Mechanical Engineers (ASME) Code, Section XI (Ref. , and is bcsed on the need to perform such surveillances und the conditions that apply during an outage and the potent 1 for an unplanned transient if the Surveillance were p formed with the reactor at power. g {wc.aAucd (continued) WCGS-Mark-t:p oj'NUREG-1431 - Bases 3.4 B 3.4-88 !/1/SM7
l RCS Leakage Detection Instrumentation B 3.4.15 ! BASES SURVEILLANCE SR 3.4.15l REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere part % 1ateiandfga_seou_s radioactivity monitors. The check gives reasonable confidence that the channel +e are operating properly. The Frequency of l 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. f SR 3.4.15.2 l SR 3.4.15.2 requires the performance of a COT on the required i containment atmosphere particulate and' gaseous; radioactivity t monitors. The test ensures that the monitors can perform 4ts their function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation. SR 3.4.15.3. SR 3.4.15.4. and SR 3.4.15.5 I lhese SRs require the performance of a CHANNEL CfLIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, ! including the instruments located inside containment. The l Frequency of@nths is a typical refueling cycle and considers , channel reliab ility. Again, operating experience has proven that l this Frequency is acceptable. l L redbs. {Q3.4.Gm-l { REFERENCES 1, 10 CFR 50, Appendix A, Section IV, GDC 30.
- 2. Regulatory Guide 1.45,
- 3. FSAR USAR, Section 5.2.5
- 4. NUREG-609, "Asy m etric Blowdown Loads on PWR' Primary i
Systems,"'1981. I
- 5. Generic Letter 84 04, " Safety Evaluation of Westinghouse
- Topical' Reports Dealing with Elimination of~ Postulated pipe Breaks in PWR Primary Main Loops.~
WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 98 'l/58 7
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.1-1 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-38 Comment: .TSTF-105 has been rejected by the NRC. FLOG RESPONSE: The July 27,1998 indust t ? s reler status reports indicate the status of TSTF-105 as rejected by the NRC with the TSTF considering. The FLOG has reviewed the traveler and is withdrawing the traveler from the conversion application. For Diablo Canyon, the CTS will be used which does not require a specific method for measuring RCS flow. This difference from the STS is l justified by revised JFD 3.4-38. ATTACHED PAGES: Attachment 8 CTS 3/4.2 ITS 3.2 Encl. 2 2-15 Encl. 3A 8 Encl. 3B 7 Attachment 10 CTS 3/4.4 ITS 3.4 Encl. SA Traveler Status page,3.4-4 Encl. 5B B 3.4-7 Encl.6A 7 Encl. 6B 6 i l i i U 4 f
i I POWER DISTRIBUTION LIMITS ; 3/4.2.5 DNB PARAMETERS I LIMITING CGNDITION FOR OPERATION i ACTION- (ContinM
- 4. !f:... '; nd E-T !'70$4s-123??
4 prior to irs- 7,gTHERMAL cMho4ause ef the POWER out-of-limit above the reducedcondition THERMAL POWER ' "'#-**** u-limit required by ACTION 1.b and/or 3, above; subsequent POWER OPERATION may proceed provided that the ind'ested RCS total flow rate is demonstrated to be within the region of acceptable opersbon prior to exceeding the following THERMAL l POWER levels: '
^
- a. A nominal 50% of RATED THERMAL POWER,
- b. A nominal 75% of RATED THERMAL POWER, and j
- c. Within 24 hours of attaining greater than or equal to 95% ;
of RATED THERMAL POWER. j SURVEILLANCE REQUIREMENTS I 4.2.5.1 The provisions of Specification 4.0.4 are not applicable to Specificabon 3.2.5.c. 4.2.5.2 Each of the parameters of Table 3.21 shall be venfied to be within their limits at least once per 12 hours. DS'-It. - 4.2.5.3 The RCS total flow rate indmetors shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. l CP 3.2-00 lI I~ undelets.,- _ - . - 3
-tmde.te4s- - _
** -: 9;;l-- - ' '::1 4.2.5.4The RCS total flow rate shall _ . . .
g a 4,,i g f
-- .:n _--- _ .: ;"at least once per
'.'" l..' ::;: ggggg
_d ._ ;i. Uc_ __5 bb_bb2 55'N.b255bb. _. Ch.s=EEEeEn?.E z"R.E gg3]
.h.s-415.5 ^. ' :t:^ : r-'M 9-'! 5e !- ;::^- f ' r '-6;; :nd '-- . : --
{g}g
- -- :: ; -' '-- ' : :: p i ? - : .".:.
i i {# Not required to be M r:f until 7 days after achieving 2 96 % RTP) { { WOLF CREEK - UNIT 1 3/4 2-15 Amendment No. 61 Mark-up ofCTS3N.2 MI3/97
. . - - - - . - . - - - _ . - . - - ..._~ ... - - . - .. - - . - . - . - . _ _ . .-
k l 4 CHANGE I NUMER _EC DESCRIPTION 3,.g g l l - _ _ _ ! ! n. reposed hanges, bas n traveler ' 95 flimina~ l ' whi clarif t reductions t be comp 1 with 2 1 rs a each determina and permi chievin j equ rium c tions for suring peak factor re t '
- i nsider o be relax ns of curre equir s. s is
{becacompletion s for thes ctivitie re not specified in cu ent TS.f
^
i - n fo4 n='_j
~
iussstr sA -eu t we. u -oo s 1 05 01 LG The designation of how instrument uncertainties are treated (nominal, in the analysis, or in the development of the TS limit) is moved to the Bases. The movement of this level of detail out of the specification is consistent with NUREG 1431 l and is an example of removing unnecessary details from the TS in accordance with 10 CFR 50.36. 05 02 LS-7 Not applicable to WCGS. See Conversion Comparison Table ; (Enclosure 38). - 05 03 LG [ sten ith EG 14 , the equ ( t fo ~gP a 3.2.-oo I ] l l E I ION he R me rs t1 st nce
! 18 hs the uirene t no 11 t c nel ae f ed the B s for t R fl - r cto, tr ,
fu on in , Secti 3.3
- t. sos conver: mon chmperr..i,v, Tpv4.a (.ppuch in tauss. )
. Noto I dnc.t.sww 3 % ./
05 04 LG [ Cons with ustr raveler TSfF 105, the xplicit ^ l r ts t the flow be asured t tb use of q isio at ance measu t and t the ' i str ation ed in the rformanc f the c orimetr i ;
' flow asur t be cali ted withi a speci time riod I o rfo g the sea ement is ed to icensee ntrol documen . The requ t to erify t the R Il f1 s within its remai ' ithin Technic j
ificati . This is a example removinpunnece ary details the TS c is accep ble based on the idance ' Q in 10 CFR .36. iussser 3A-ee 05 05 LG Consistent with NUREG 1431, Wolf Creek specific REQUIRED ACTIONS would be modified to move details regarding identifying the cause of RCS low flow rate to the Bases. This is acceptable because it would remove details that'are not required to be in TS to provide operational safety while retaining the limiting conditions for operation. 05 06 LS 8 In accordance with NUREG 1431, if any of the DNB related parameters of pressure, temperature, or RCS flow are found to be outside their limits, the time period required to perform a power reduction would be extended to 6 hours. The DNB related , parameters of Reactor Coolant System (RCS) average temperature, pressurizer pressure and RCS flow rate are WCGS-Description of Changes to CTS 3M.2 8 5/15/97
I INSERT 3A-8 WC 3.2-001 l 04-11 A A note is added to Wolf Creek CTS SR 4.2.4.2 to indicate that the surveillance'is not required to be performed until 12 hours after input from one Power Range Neutron Flux channel is inoperable with THERMAL POWER > 75% RTP. This change is considered an administrative change since Action A.2 provides a frequency of 12 hours to determine OPTR when OPTR has exceed 1.02. Further justification is based on the fact that under normal circumstances, OPTR would not be expected to change significantly l within a 12 hour period. If a significant change in OPTR were to i occur, it would likely be the result of control rod misalignment which would likely be detectable immediately by means of the rod deviation monitor or rod bottom lights. Additionally, a note is added to CTS SR 4.2.4.1 to indicate that CTS SR 4.2.4.2 may be performed in lieu of this surveillance requirement to confirm the indication of the remaining three ! excore channels. As identified in the NUREG-1431 Rev. 1 I definition of QPTR and the SR 3.2.4.2 Bases, OPTR is a ratio of I excore detector outputs and, for the purposes of monitoring the OPTR when one power range channel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated OPTR and any previous data indicating tilt. INSERT 3A-8a 0 3.4.1-1 CTS SR 4.2.5.4 provides descriptive detail of the method for the determination of RCS total flow rate during a Surveillance. This detail is moved to the ITS 3R 3.4.1.4 Bases. These details are not necessary to ensure the RCS total flow rate is within required limits. The requirements of ITS SR 3.4.1.4 are adequate for ensuring the RCS total flow rate is within required limits. These details are not necessary to be in the TS to ensure the RCS total flow rate is within required limits. Moving this information maintains consistency with NUREG-1431. Any change to this descriptive information will be made in accordance with the Bases Control Program descrit ed in ITS Section 5.5.14. l l
CONVERSION CO'iPARISON TABLE - CURRENT TS 3/4.2 Page 7 of 8 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY (4-Il A luSERil 35 I J-- -[ iJc. 3.2.-00 - - - 3( - -
, _ -~ _ _
pre 4A - j 04-10 The al ti or t requir t to set t ower pr- WA Jae" WA N6 *JA \ 1Er14 Ra Neutro ux- h set nt dur power educt ~ y ired QP TIONS ld xte o 72 rs.fdr '- k o.3.2. ] k.
' Wolf Cr t44(1 sed.
05-01 The designation of how instrument uncertainties are treated Yes Yes Yes Yes LG (nominal, in the analysis, or in the development of the CTS limit) is moved to the Bases. 05-02 The CPSES specific requirement to verify that the total RCS No Yes No No LS-7 flow is within limits using the plant computer or elbow tap output voltage on a monthly basis is deleted. 05 03 f The requirement t: - rf:= C:'ffr'. C'i!='.^7F=> :t 1::;; yes No-N; .h CTS Yes Ves.4.-Ost A LTs Ves >5.- dat.a cT:s LG =:: per 10 .~4h; e,-J L .mp . .. .a to normalize the 8CS loopfle a ute "mJ 3c-dor.s. is th;--,c? r: moved tc R.e Bases for the surveillance _ requirements for tk 200 != r ::ter trip f:r tion 7 g p 3,g %] s ( 4e ITS 4-3 2.+.I _ 05-04 'Consis~ nt h1~ ' try t eier -105, exp cit No - Requirement Yes Yes Yes LG r ts th the R flow asur the the se q not in CTS. , __ _
-4he fim ,{ c2%r2}ian a ecisi heat ance me uremen and th the cal pg g,.g.A yq,rj l1 rumen ion us in the rforma e of metr _ ,
af 4he. Asiriament26on used in the perft>ymuca_ _ 43,4 ~ g , y {l low me uremen calib- ted wi nas ift time of the. calorim etr'ic. Oo o measu.re m e n t is y gi of perf ing t measur _ _n t is v_d e _ the ases. mg .to h Bases. .s-The Wolf Creek required actions would be modified to move No No Yes No 05-05 LG details regarding identification of the cause for low RCS flow rate to the Bases. The time to reduce power to less than St RTP would be Yes Yes Yes Yes 05-06 LS-8 revised from within 4 hours to within the next 6 hours. 05-07 This surveillance is modified to require that it be No - See ITS No - See Yes Yes H performed within 7 days of achieving 95% RTP. Section 3.4. CN CN 05-11-A. 3.4-51. WCGS-C<mrersion Comparison Table - CTS 3M.2 5/15/97
~. -_. _- - . _ -
4 i INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 TR._VELER # STATUS IIIFFERENCE # COMMENTS TSTF-26 Incorporated 3.4 32 Approved by NRC. TSTF-27, Revh Incorporated 3.4-33 (dpproM MhQ 3.4~2 -Il TSTF 28 Incorporated 3.4-22 Approved by NRC. TSTF-54, Rev. I Incorporated NA (Ahp N 8 JAM *T8L.5.&m1l TSTF-60 Incorporated 3.4-15 Approved by NRC. TSTF-61 Not Incorporated Minor change that is adequately addressed in the Bases. l TSTF-87, Rev.h incorporated 3.4-31 {Qe5M bj AlQ /TA3.4mXHA J
~
TSTF-93 Q Incorporated 3.4-17 {Appr M 6[ dR Q /93.4. ?JJ TSTF-94 h Not Incorporated NA Retained current TS. [TM 3.y-co s-[ g0V22_InMorat4_/ #4J36 'Q 3.4.b Ll TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply. TSTF-il3, Revh Incorporated 3.4-39 Q 3.4.11-3} TSTF-il4 Incorporated NA Approved by NRC. TSTF-Il6, Revh' incorporated 3.4-36 l 4.3.4. 83-2J TSTF-136 Incorporated NA ( kg M/(C] /pt.3.9oot f TSTF-137 Incorporated NA @ppe,vsd. by MMJ/74 3.4- So ? l TSTF-138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6.. TSTF-151 Q incorporated NA /T4 J.&ootj TSTF-153 Incorporated 3.4-01 $M by A/Ac]/723.5-aop} TSTF-162 incorporated NA (A prawA n j NAC.]/7E J.4- soc,} D O G M E ef" D incorporated M3.4-45%@ See also Cns 3.4-18 and 3.4-20.& 2.4.81-L T L M ~7srF *2A ) Incorporated 3.4-35 [4 3.4tl.2.1 DbOC.WRawO Incorporated 3.4-10 DCPP onlykPra b %I787 '-oof I ([WOU-87,lav) Incorporated 3.4-47 [ Q 3.4.88-4 ) M9 Incorporated 3.4-40 Applicable to Callaway and Wolf Creek only. [Q 3.4.8-2.] (TsTs:- 28Q
'29]0 Incorpora:ed 3.4-49 \ Q J.+.I'l- I l S/15/97
RCS Pressura, Temperatura, and Flow DNB Lizits 3.4.1 1 SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is a E22001 12 hours $$$PS::'d 2220 psig, l SR 3.4.1.2 Verify RCS average temperature is s-E5013 12 hours $t84PShfh i 590J5'F. I 1 l SR 3.4.1.3 Verify RCS total flow rate is a [204.000] 37.71 ' 12 hours ; xgD'gpe. liBBFPS$~W l 1 SR 3.4.1.4 - -..--
-NOTE - - -- - -
, 1 Not required to be partermed until 24 hours ! M
7Idays after a 95 % RTP. i/3.4-40%
. . .g .g.
VerifyMFcchjerTbc;t b ler.hc S. t6teet(swed 18 months gg ;g;pggg RCS total flow rate is a 1204.000] 37.1.x'10' . gpm. P IBf! M WCGS-Mark-up ofNUREG-1431 -ITS3.4 3.4 4 $/1587
e ! RCS Pressure. Temperature and Flow DNB Limits !- B 3.4.1 1
- - BASES SURVEILLANCE SR 3.4.1.3
! REQUIREMENTS (continued) The 12 hour Surveillance Frequency for RCS total flow rate is perfonned
- using the installed flow instrumentation. TlHQinstaM
- i.hstrtMintatidnJproviMindithtior@[aj{[ertentagdofIMRitiiMtie j
bHsedJon1the% rec 1MonfcaToirimetric?heattbalance19? nutJ *pBLli@75 i specifhWiMwar*=gefof(the# tota.1]flowirateegulred]to. gh ;05 totallflow;fatufflimitOhs 12 hour interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions. SR 3.4.1.4 k Heasurement of RCS total flow rate' - '- ^ ^ ' 2 ;;lder. D 3 4 I-0
;;jrfi;;;ri;I; t betengponce every 18 monthsia1"tergeaciliTefuelipg
~a llows the installed RCS flow instrumentation to be celibr;ted normali' zed and verifies the actual RCS flow rate is creater than or ,
- " l to the minimum reouired RCS flow rate.1Thi~ Twi G E M4 81 a4prWcisjerf ca'prfmetrYhegt4al ~~ 4unertyperronningia precision neat (Dalance, theinstrumentat1c.n usedgfor;detetistrrtig; steam pressure.ifeedwater! pressure,. fiedwater temperatereEandifeldduater venturijaplin;thescalorimetriefcalculations shall:be;calibretafwithin 7.~daysTprior to! performing 1the; heat: balance.
The Frequency of 18 months reflects the importance of verifying flow after a refueling outage when the core has been altered, which may have caused an alteration of flow resistance. This SR is modified by a Note that allows entry into H00E 1. without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 h;urs 7 days after :: 96-95% RTP. This exception is appropriate since the heat balance requirn the plant to be at a minimum of 90-95% RTP to obtain the stated RCS fiow accuracies andithestest is only,a confirmation of:SRf3.'414. The Surveillance sFall be performed within 24 hours 77 days after reaching M-95% RTP. REFERENCES 1. FSAlHJSAR,-Seetier [15] Chapter 15. WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 7 5/1/5 M
. .- _ _. _.. - __ _ _ _ _ _ _ _ - _ _ - _~
- .. .. - ~- .-. - . . . - - . _ - - . - - - . - . - ..-. - - - . - . - -.. - .
CHANGE NIM ER JUSTIFICAT10N purposes (per Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR %ted exception) would not allow the SR to be ! considered to be met until the appropriate conditions were l available for it to be performed without entering the actions. The Note to these SRs would allow startup in Mode 3 if the SR had not been performed during the required frequency, but would limit the exception to prior I to entering Mode 2. The change is consister.t with trsveler @@DJJTF -@{q.:s.4.i n -s. } , '3.4 36 SR 3.4.13.1 and LCO 3.4.15 are revised per traveler TSTF-116. The note addresses the concern that an RCS water inventory balance connot be meaningfully performed unless the unit is operating at or near steady state conditions. The note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be a deferred for 12 hours after re establishing steady state conditions. 3.4 37 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). p 3,g,g_g1 3.4 38 r--P' Consis nt T 105 the d ails o 6d are v rified rethemet)fr
~
whi t CS f rat mov ~ 8h PP U Y' '
. 1. to t Bases Movin this i formati t t l S** 0^M*" N#" Bas , al s the se of cisio at b nc ,
(.Endeswre. ' ABS ps, othe ccepta e met s in o er c
~
th verifi ion and s cons tent w ht N Gd431 p ilosop of movi clari ng inf a njnd ive detai 5 out the T to Bases. (descr , 3.4 39 The shutdown requirements of ITS 3.4.11 would require the plant to reduce T,,, to <500*F within 12 hours, rather than MODE 4. to address the concern of entering [LTOP] LCO
=
3.4.12 Applicability with inoperable PORVs., Sr (414.H-b] m**> consistency, the shutdown requirements of ITS 3.4.16 are also revised to allow 12 hours to reduce T,,, to <500*F. This change is consistent with TS F 113. 3.4 40 Consistent with traveler . the Note to SR 3.4.1.4 l034.1-2.1 would be modified to provi e additional time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and additional time normally would be required to establish WCGS-DifferencesfromNUREG-1431-ITS3.4 7 S/15M7
Page 6 of 8 CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 SECTION 3.4 UlFFERENCE FROM NUREG-1431 APPLICABILITY DIABLO CANYON C0KANCHE PEAK WOLF CREEK CALLAWAY NUMBER LESCRIPTION r-- oc w c,.,,;. tost- de ib e.# CTS SR 4.2.3.3,the detAls en h A+ boa] whick] the. Ar.s f top c2te io veeh4 si_ n:vvso_% fvam sTs sg.4.i .4.+eS+., J 3.4-38 N;lConsis efft wit T. t detat on/he t by Yes hdo {G3 4,1-l) f . whi- the RGS flow r e are erif' arjrmo~ SR3f.1.4to I Ba .f-Yes Yes Yes Yes 3.4-39 The shutdown requirements of ITS 3.4.11 would require the plant reduce Tmto <500 F within 12 hours, rather than MODE 4. to address the concerr of entering [LTOP] LCO 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are also revised to all 12 hours to reducemT to
<500"F. This change is consistent with TSTF 113.
No - See CN No - See CN Yes I Yes 3.4 40 Yhe Note to SR 3.4.1.4 would be modified to specify a 3.4-51 3.4-34 plant specific reactor power and to provide additional time to perform an RCS precision flow rate measure n t. No No No ' LCO 3.4.1 is revised to reference Tables 3.4.1-1 and Tes - Allowance 3.4-41 added per 3.4.1-2 for RCS total flow rate limits for DCPP Units I and 2 respectively. Amerdent 60/59. No No An exception to SR 3.4.14.1 frequency to leak test Yes - Specific No 3.4-42 to DCPP PIVs 8802A. 88028 and 8703 has been added. This change is consistent with the DCPP current TS. No Yes No No 3.4-43 A new Condition is adoed to LCO 3.4.1 to reflect the current TS of Wolf C:eek for RCS Flow Rate. No Jo- Yes @3D7-3.4-44 Steam generator levels for MODES 3. 4 and 5 are specified to ensure SG tubes are covered. The No g @D _6
)
Epa 9djuc_currentTSdidnotensuretubecoverage. 5/1587 WCGS-Conversion Congparison Table-ITS3.4 h s.__ .-_u__ _ . _ _ _ . . _ - _ _ _ . . _ _ _ . - _ _ _ _ - . _ _ - . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ -
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.1-2 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-40 Comment: WOG-99 has not yet become a TSTF. FLOG RESPONSE: WOG-99 has been designated TSTF-232 which is currently under NRC review. No changes to the ITS mark-ups were made in the process of assigning this traveler a TSTF number. As explained in Enclosure 6B to Attachment 10, JFD 3.4-40 does not apply to CPSES or DCPP. Those plants are retaining their CTS, as explained under JFDs 3.4-34 and 3.4-51, respectively. Callaway and Wolf Creek continue to pursue the changes proposed by this traveler. ATTACHED PAGES: Attachment 8, CTS 3/4.2 - ITS 3.2 Encl.3A 9 Attachment 10, CTS 3/4/4 - ITS 3.4 Encl. 5A Traveler Status page Encl.6A 7
1 I CHANGE NUISER lGC DESCRIPTION 1 i maiQined within specified limits in order to ensure j consistency with the assimed initial conditions of the i accident analyses. The limits placed on the RCS temperature. l pressure, and flow ensure that the minimum departure from Nucleate Boiling ratio (DNBR) will be met for each of the transients analyzed. Compliance with the above limits is verified every 12 hours. If a parameter is found to be outside the required limit, 2 hours are allowed in order to restore the parameter to within the limit. If the parameter is not restored to compliance within the required time the plant must be shut down. The revised completion time of 6 hours is acceptable to allow transition to the required plant conditions in an orderly manner without unnecessarily initiating any undue plant transients and on the small likelihood of a severe event occurring during the extended 1 1 time period. J I 05 07 H This surveillance for measuring RCS flow by precision heat i balance is modified to add a footnote that corresponds to the ) i Note for ITS SR 3.4.1.4. The footnote requires that the j surveillance be performed within 7 days of achieving 95% RTP. l This is more restrictive in that it ties the surveillance to the beginning of a cycle. This is acceptable because other ! indication of RCS flow is available (RCS flow meters) and time is provided to establish This plant conditions is consistent suitable for the@ with traveler W- 2.s krecision heat In addition, balance. the THERMAL POWER specified in the footnote would be changed from the generic value provided in NUREG 1431 g -24 to a plant-specific value of x 95 % RTP. This change is 3 acceptable because it specifies a THERMAL POWER in better 2 agreement with current operating procedures for performing a precision heat balance. Current TS do not'specify a power j level for this measurement.
]
05 08 Not used. I l 05 09 LG The requirements for inspecting and cleaning the feedwater flow venturi would be moved to licensee controlled documents..
! These details are not contained in NUREG 1431. This is an example of moving unnecessary detailed information from the TS and is acceptable.
05 10 A Not applicable to WCGS. See Conversion Comparison Table ; (Enclosure 38). ) i t 05 11 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B).
- 2. A INSERT 34 [- h [ GP 3.~2. -00 I \
WCGS-Description ofChanges to CTS 3M.2 9 5/1587
4 J i INDUE,1RY TRAVELERS APPLICABLE TO SECTION 3.4 TR AVELER # STATUS DIFFERENCE # COMMENTS f i TSTF-26 . Incorporated 3.4-32 Approved by NRC. TSTF-27 Revh incorporated 3.4 33 [Aggravd (Q 3.4.'2.-Il l TSTF 28 incorporated 3.4 22 Approved by NRC. I _ _
- TSTF 54, Rev.1 Incorporated NA @p[ h *MM*ns..I *m9l l TSTF-60 Incorporated 3.4-15 Approved by NRC.
I j TSTF-61 Not Incorporated Minor change that is adequately addressed in the
- Esses.
TSTF-87, Rev.h incorporated 3.4-31 (dppieywJ.,63 M Q /nf.3.4924I TSTF-93 @ Incorporated 3.4 17 {Aprreved 6[ hA Q /93.4. iJ l TSTF-94 h Not incorporated NA Retaind current TS. [TM 3.v-wsf l [TS p 0 V f fin 11whoratC _y . - __ 14 3.4.F i1 f TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply, f TSTF-ll3, RevhI Incorporated 3.4-39 4 3.4.Ii-3) TSTF-ll4 Incorporated NA Approved by NRC. TSTF-116 Rev[h Incorporated 3.4 36 I4J 4.83-1] j TSTF 136 Incorporated NA (APPr'VO kS h M c ] /7:n.3.6 cot / TSTF-137 incorporated NA [4ppewvsd. 6g #MJ/74 3.4- Sof f .l TSTF-138 Not incorporated NA Inccasistent with RCS loops ; requirements of ITS 3.4.5 and j 3.4.6 i TSTF-151 Q incorporated NA /T4 J.e do*l l TSTF-153 Incorporated 3.4-01 hrwvv4., by NED /nt J.9-nog
. _ _ - - l TSTF 162 Incorporated NA (ApprM by NK.lfME,J.4- 56) l W T) See also Cns 3.4-18 and 3.4-20.19 8481~9 Incorporated h3.4-45%5%
(M "T51'Fa$ Incorporated 3.4-35 [43.4.al.t( . hWOC#:"Rawr0 Incorporated 3.4-10 DCPP onl(4 P E P N I*f NR".)MJ Mf I i (WOd-87 Rev) Incorporated 3.4-47 [ 4 3. 4.81- 4 ) M" incorporated 3.4-40 Applicable to Callaway and I (TisTF-2sg Wolf Creek only. [@ 3.4.8-2.] h Incorporated 3.4-49 M.3.+. L'2. - 1 i 0 5/15/97
CHANGE NUMER JUSTIFICATION l purposes (per Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR Noted exception) would not allow the SR to be considered to be met until the appropriate conditions were available for it to be performed without entering the actions. The Note to these SRs wopld allow startup in l Mode 3 if the SR had not been performed during the l required frequency, but would limit the exception to prior to entering Mode 2. The change is consistent with traveler @E3DJJTF-@ q.3 4.o 2 ] 3.4 36 SR 3.4.13.1 and LC0 3.4.15 are revised per traveler TSTF-116. The note addresses the concern that an RCS water inventory balance connot be meaningfully performed unless
- the unit is operating at or near steady state conditions.
The note adoed to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be ! deferred for 12 hours after re establishing steady state conditions, f j 3.4 37 Not applicable to WCGS. See Conversion O rison Table (Enclosure 3B). yq3,,gg_g\ 3.4 38 __ g---P'"Consi s nt hT 105 the ails o the met T whi t CS f rat are v rified re move fr ! *f *Pf#' *
- 1. to t Bases Movin this i formati t t S** G e** "" Y
- M
- l L
Bas , al s the se of cisio at b nc ,
'ge (Enc.loswe. ~6BT /
' ~
ps, other ccepta e met .s in o er fo th verifi ion and s cons tent w ht N EG- 431
'p ilosop of movi clarip, ng inf a n pd Qeser ive detai s out of the T to Bases. ,
l 3.4-39 The shutdown requirements of ITS 3.4.11 would require the plant to reduce T.., to <500 F within 12 hours, rather than H0DE 4, to address the concern of entering [LTOP] LC0 3.4.12 Applicability with inoperable PORVs., For
' ""% > consistency. the shutdown requirements of ITS 3.4.16 are 14A S.H-b]
also revised to allow 12 hours to reduce T,,, to <500*F. This change is consistent with ! 3.4 40 Consistent with traveler . . the Note to SR 3.4.1.4 l410-2.1 ! would be modified to provi e additional time to perform an i RCS precision flow rate measurement. The time allowed
- would be changed from 24 hours to 7 days. This change is
( acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and j additional time normally would be required to establish , WCGS-Differencesfrom NUREG-1431 - ITS 3.4 7 3/15/97 I _, _ _ . - _. __ - _ _ . _ __ - _ _ _ ___ .. --
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.2-1 APPLICABILITY: DC, CP, WC, CA REQUEST: Difference 3.4-33 Comment: TSTF-27 Rev. 3 is still pending NRC approval FLOG RESPONSE: The July 27,1998 industry traveler status reports indicate the status, of TSTF-27, Rev. 3 as approved by the NRC. The proposed wording in TSTF-27, Rev. 3 was modified from TSTF-27, Rev. 2, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes approved in TSTF-27, Rev. 3. ATTACHED PAGES: Encl. 5A Traveler Status page Encl. 5B B 3.4-10 Encl.6A 6 i I i
i INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.4 i 1 TRAVELER # STATUS - DIFFERENCE # COMMENTS l TSTF Incorporated 3.4-32 Approved by NRC. TSTF-27, Revh Incorporated 3.4-33 QM 14 Q Q 3 4.*L-il 4 TSTF-28 incorporated 3.4-22 Approved by NRC. TSTF-54, Rev. Incorporated NA (4_ph@_'k 4%Fla..t+eo91 TSTF-60 Incorporated 3.4-15 Ar.oroved by NRC. s
- TSTF 61 Not incorporated Miaor change that is I
+ adequately addressed in the I 4 Bases. I TSTF-87, Rev.h incorporated 3.4-31 (ippd. m M /FW.3.4924l e l TSTF 93 @ Incorporated 3.4-17 (Apreend 6 NAQ /93.4. i3l _ l TSTF-94 h_ -_ -__ Not incorporated NA Retained cairrent TS. [TM 2.4-desf - l
- [TS f 0 V f fInr.pf5eratpr_ y e l4 3.4.E t I' TSTF-108, Rev. I Not incorporated NA LCO 3.4.19 does not apply.
TSTF-ll3, Revh ! Incorporated 3 4-39 4 3.4.ll-3) 4 i TSTF-ild Incorporated NA Approved by NRC. f TSTF-116. Rev[' Incorporated 3.4-36 l 4.3.4. G -d TSTF-136 incorporated NA (Afr"WAk5 P N a c D /71s.3.4oot / i TSTF-137 incorporated NA [4pp md. 63 #M37A 3.4- $89 I !- TSTF-138 Not incorporated NA Inconsistent with RCS loops l requirements of ITS 3.4.5 and 3.4.6.. l , TSTF-151 h Incorporated NA /T4 J.(Hotl TSTF-153 Incorporated 3.4-01 rwed, by AiAQ /nt 2.9-noQ TSTF 162 Incorporated NA (kpprandbyNAC.)/7%.J4** G-} GEOG 4t3ef'D Incorporated M 3.4-4 % S )2. See also Cns 3.4-18 and 3.4-20.I9 8.4 3L*1_I (W(A40T5fF"2JS Incorporated 3.4-35 [4 3.4.61 -1.\ BWOG.69: RawO Incorporated 3.4-10 DCPP onlyhy; AM h IRE ~)NJ.Nf I ({WOd-87,'Rev) Incorporated 3.4-47 [4 3.4.81-4j M* Incorporated 3.4-40 Applicable to Callaway and (Tarts:- 2dllQ Wolf Creek only. {c 3.4_.$ Incorporated 3.4-49 % 3.+.t1- 1 { S/158 7
RCS Ministan Temperature for Criticality ; B 3.4.2 i BASES 1 1 1 i 1 ACTIONS L J (continued) 1 i ritii$FaimentsjditNihithiRalTowed!C6mp1bteniTiimeEintatdiiWiFiedliaMil 4 PERl8BiBoC180erisRttinMatl1s[643ojiUMMB3lB! 4 8LIMlW2cotiidttid gFpaydt64tegLuptiot([oggyplenJlgl Rapid i i reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating j' experience, to reach MODE 3 in an orderly manner and without chal L .ging plant systems. I SURVEILLANCE SR 3.4.2.1
- REQUIREMENiS i
RCS loop average temperature is required to be verified at or above 5+1
- 551'F every 30 ;;;inutesThours &n [T 4, dedtien, l= l= T,,,3 l
cl: = =t re::t :~' :ny ECS 1:0p T,,,em7 [g ]*F.- i I ?.; %et; aedific ; th; SR. 'J.ca eny RCS ia;; ;;;r;;; t@cr;tur; i; { < [517]*F :nd th [T,,, T demti:n l= != T ,,) ;l;r;;; i; ;l;r.;.ing, RCS 14 ;V;r;;; t47;tur;; ;;uld f;11 bel;W tt~ LOO r;quir; at____ 1 ' WitP;;t ;dditica;.1.;; rains. The SR to verify operating RCS loop W342-l] aver?ge temperatures every 30 ;;;inut;; 12: hours)1sjeequent enougrw;
@rgarit thodnadvetenVviolation4f ttrfCD/ana: takes;1ntoyaocount 1naications:and: alarms thattare continuously;available to;tteoperator intthesontrol~roomcfMicon.s. ten + sim wr routim4. ' ~
(Surwlillmcas tahich are.IyricM3 perfbrmai oncA Per shin. 6 REFERENCES 1. FSAR USAR. Secti;n [15.0.3] Chapter 15. i in didn, operabs we. irainad tb E4. seMtse.t OIW"'I 4emperabe. du.rQ approach b c/ihcAfih u4. usilenwe. 4 hat 4he. rnihiMum tempesettu.re er c.rWicAlitq *is me.t as crWi c2hh th approat.hed. ___ WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 10 5/1/SM7
CHANGE l' NLMIER JUSTIFICATION increasing MODE would reduce the risk of a low temperature overpressurization event. In these cases it would be unwise to maintain the plant in a lower MODE configuration. Increasing plant MODE may also be the expedient way to exit a low temperature overpressurization potential when operating within a CONDITION. This option should be retained as exists in the current Technical Specifications. 3.4 31 The ACTIONS in ITS 3.4.5 and 3.4.9 are modified to reflect their LCO. The position of the reactor trip breakers and the power supply status of the CRDHs are not LCO requirements: therefore, the CONDITIONS and ACTIONS are revised. As worded in NUREG 1431 Rev.1. these ACTIONS could preclude certain testing in H0DE 3. A more generic l action, which assures rods cannot be withdrawn, replaces these specific methods of precluding rod withdrawal. The l specific methods are added to the Bases as examples. The l revised ACTIONS still assure rod withdrawal is precluded l and this detail is not required to be in the TS to provide l adequate protection to the public health and safety. No l~ technical changes result from this change. These changes i are consistent with traveler TSTF 87 Rev.1. I l 3.4 32 In accordance with traveler TSTF 26 the ACTION would be
- changed to specify taking the plant to a H0DE for which l the LCO is not applicable. This change maintains the
! consistency between the Mode of Applicability and the Required Action which requires the Mode of Applicability to be exited. l 3.4 33 The Frequency of SR 3.4.2.1 to verify operating RCS loop average temperature at or above [551]' F is changed to 12 hours from the current surveillance frequency of 30 minutes. The SR to verify throperating loop average temperatures every 12 hour suffi ntly equent toJ q 3, q ,3 i najysrtTnt vio on of _ LC nfdonsiders s
@inalcations and alarms tnat are continuously available l
the operator in the control room. This change is based on j industry traveler TSTF 27. 3.4 34 Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 38). i [ 3.4 35 This change adds a note to SR 3.4.11.1 and SR 3.4.11.2 ( stating that the SRs are only required to be performed in { Hodes 1 and 2. The Actions Note "LC0 3.0.4 is not
- applicable" is intended to allow Mode changes for testing WCGS-DifferencesfromNUREG-1431-ITS3.4 6 $/1SM7 4
- _ _. . = - -_ - _ . _ _ . - . ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.3-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.4.3 Bases References Comment: WCAP-14043-NP-A, Rev. 2 January 1996, has replaced WCAP-7924-A, ) April 1975. Please summarize the differences / applicability to the FLOG. l l FLOG RESPONSE: WCAP-14040-NP-A, Rev.1 was NRC approved as an acceptable l reference for " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Curves" by SER dated 10/16/95 with minor comments which did not affect the SER. These comments were incorporated and the WCAP-14040-NP-A was issued as Revision 2 in January 1996. NRC acceptance of this WCAP as a reference was based upon the following key elements:
- 1) The WCAP incorporates state of the art fast neutron radiation j
transport.
- 2) The WCAP cold overpressure mitigating system satisfies SRP Section 5.2.2 and BTP RSB 5-2.
- 3) The WCAP fracture mechanics calculation conforms to 10CFR50, Appendix G and SRP Section 5.3.2.
- 4) The WCAP conforms to Reg. Guide 1.99, Rev. 2 in calculation of the adjusted reference temperature.
- 5) The WCAP conforms to 10CFR50, Appendix G for methodology i for calculating minimum temperature in the P-T limit curves.
- 6) The WCAP satisfies the provisions of the draft generic letter published in the Federal Register for comment of June 2,1995.
These items are consistent with the STS reviewer's Note on STS 5.6.6. Plant Specific Discussion: Wolf Creek removed surveillance specimen capsule V during the ninth refueling outage in November 1997. This capsule is currently being evaluated, and the methodology of WCAP-14040-NP-A, Rev.1 is being utilized for the Cold Overpressure Mitigating System setpoint and heatup and cooldown curve development. The Withdrav.n Specimen Test Results Report will be submitted to the NRC by the end of September 1998. Therefore, Wolf Creek believes it is acceptable to reference WCAP-14040-NP-A, since the WCAP has been approved by the NRC, and Wolf Creek is following the methodology provided in the WCAP. ATTACHED PAGES: None
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.-4-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.4.4 Bases Comment: The Bases refer to the DNBR limit in the safety limits. Where is it? (this appears to be a problem with the STS, as well as these conversions). FLOG RESPONSE: As described in the Applicable Safety Analyses Bases for ITS Section 2.1.1, the DNBR limit is: "There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB." The actual numerical value is specific to a given DNBR correlation and analytical methodology. The correlations and methodologies are NRC-approv.ed. More than one correlation or methodology, as generally documented in the USAR, may be used depending on core design and the particular transient being analyzed. For this reason, a more general term such as "DNBR limit" is used. This convention has been used throughout the Bases for ITS Sections 2.0, S 1, 3.2,3.3, and elsewhere in 3.4. In the process of responding to this RAI, it was noted that all FLOG plants except DCPP and CPSES have a markup methodology error in the second to last paragraph of the Applicable Safety Analyses Bases for ITS Section 3.4.4. The acronym "SL" should have been struck-through; this is addressed under Comment Number 3.4. Gen-1. , ATTACHED PAGES: None 1
ADDITIONAL INFORMATION COVER SHEET , i ADDITIONAL INFORMATION NO: Q 3.4.5-1 APPLICABILITY: WC, CA REQUEST: Change 1-14 LS-22. (Callaway and Wolf Creek) i Comment: The change discussion is not adequate. The NSHC contains the necessary justification. FLOG RESPONSE: DOC 1-14-LS-22 is revised to read:
"The LCO and ACTION b of Specification 3.4.1.2," Reactor Coolant System, Hot Standby," would be revised to require that two reactor l
coolant loops be OPERABLE. Loop operation requirements would also l be revised to be contingent on Rod Control System status. The requirement to have a third OPERABLE reactor coolant loop would be deleted, consistent with NUREG-1431. This is acceptable because the MODE 3 decay heat removal requirements are sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay i heat. A second RCS toop ensures redundant capability for decay heat l removal. When the Rod Control System is capable of rod withdrawal, two I loops must be in operation to er.sure accident analysis assumptions are satisfied. When rod withdrawal is precluded, only one loop is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident analyses which assume only two RCS loops in operation include the Uncontrolled RCCA Bank Withdrawal from Subcritical and the hot ' zero power RCCA ejection events. The initial ccnditions and analysis assumptions for these events will be unchanged since two loops must still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal. These reactivity transients rely on the Nuclear instrumentation System's high flux trips for event termination which occurs very rapidly (on the order of seconds). There would be no benefit of having a third RCS loop OPERABLE for these transients since by the time the loop could be brought into operation, the event would be over for all practical purposes." ATTACHED PAGES: Encl. 3A 3
L
=
_ CHANGE Nl#EER IL9]C DESCRIPTION 1 11 M This change adds a new surveillance for verification of breaker alignment and power availability to the required pump not in' operation. This change is in conformance with NUREG 1431 Rev. 1.
- 1 12 M The Actions are changed to separate the required act wns f for only one required RHR loop OPERABLE and no required L RHR loops OPERABLE. These revised Actions are consistent with the Actions which are required under this LC0 in NilREG 1431 Rev. 1, and are more conservative than current required actions.
1 13 M Not applicable to WCGS. See Conversion Comparison Table Z (Enclosure 3B). 17
~
1 14 LS 22 The L and ion are r sed ' reg re o yt OP LE th t loops 'n oper ion " n t ro con ol e yst s ca le of od wit rawal do lo in / [ o ation n th rod cogitol sy em i not pa e o[ , - od wit rawal This change i onsi ent ith REGE1431) Rev. I N$ EAT 3A -3 gg 1 15 M A steam generator (SG) level corresponding to 10% of the wide range does not cover all of the SG tubes. To qualify = as a valid heat sink, the tubes must be covered. This is z a more restrictive change. [,] s. +. p 1 16 A Consistent with the intent of traveler TSTF 153, this change revises the note that permits up to I hour "deenergization" of RCP/RHR pumps. The revised wording r clarifies the intent of the note to allow the pumps to be
" removed from operation" instead of "deenergized", thus permitting other means of removing the pumps from service.
__ With this change the pumps are not reauired to be deenergized to use the note (e.g. the pumps may be isolated, etc.). The change is considered to be , administrative because from the standpoint of providing an exception to the LC0 requirements (to maintain the operability and operation of the pumps), the revised wording is equivalent. g 1 17 LG Not applicable to WCGS. See conversion Comparison Table 7 (Enclosure 38). L g WCGS-Description of Changes to CTS 3M.4 3 S/15/97 =
INSERT 3A-3a 0 3.4.5-1 l The LC0 and ACTION b of Specification 3.4.1.2, " Reactor Coolant System, Hot Standby," would be revised to require that two reactor coolant loops be OPE RAB L E. Loop operation requirements would also be revised to be contingent on Rod Control System status. The requirement to have a third OPERABLE reactor coolant loop would be deleted, consistent with NUREG-1431. This is acceptable because the MODE 3 decay heat removal requirements are sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay heat. A second RCS loop ensures redundant capability for decay heat removal. When the Rod Control System is capable of rod withdrawal, two loops must be in operation to ensure accident analysis assumptions are satisfied. I When rod withdrawal is precluded, only one loop is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident analyses which assume only two RC3 loops in operation include the Uncontrolled RCCA Bank Withdrawal from Subcritical and the hot zero power RCCA ejection events. The initial conditions and analysis assumptions for these events will be unchanged si.nce two loops must still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal. These reactivity transients rely on the Nuclear Instrumentation System's high flux trips for event termination which occurs very rapidly (on the order of seconds). There would be no benefit of having a third RCS loop OPERABLE for these transients since by the time the , loop could be brought into operation, the event would be over for all I practical purposes. INSERT 3A-3b 0 3.4.5-2 l 0 3.4.5-3 Six percent of the narrow range span is specified at the higher temperatures of MODES 3 and 4 whereas 66% of the wide range span is specified for MODE 5. , ! Both values ensure SG tubes are covered. The Emergency Operating Procedures l cite the 6% narrow range level to ensure heat sink adequacy. l l l' i I i
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.5-2 APPLICABILITY: CA, WC REQUEST: ITS SR 3.4.5.2 (also SR 3.4.6.2 and SR 3.4.7.2) (Callaway) Change 1-15M Comment: The sections of the ITS use the phrase "or equivalent" yet the term is not explained in the change or in t- ITS Bases. According to the information provided narrow range level is used at tt higher temperatures (Modes 3 and 4) and wide range level is used at the lower temperatures (Mode 5). If "or equivalent" means using the wide range at higher temperatures and the narrow range at lower temperatures are the levels specified applicable at the different temperatures? If not, what are the equivalent levels to the values specified in the ITS and how were they determined? FLOG RESPONSE: At Callaway, the top of the highest steam generator (SG) tube is 344 inches above the tube sheet. The wide range instrumentation provides level indication from 7 inches above the tube sheet (0% indication) to the moisture separators (a range of 559 inches). The narrow range instrumentation provides level indication between 438 and 566 inches above the tube sheet for its 0-100% indication (the use of a common upper tap results in 100% level indication on both wide range and narrow range nominally being at the same 566 inches above the tube shaet). A calculation was puformed to correlate the top of the highest tube to the wide range scale for MODE 5 conditions (the wide range instrumentation is calibrated for cold conditions), with margins added in for instrument loop errors and readability, resulting in the specified 67% wide range level. A minor error in the calculation was corrected, resulting in the specified 66% wide range level value cited in the attached pages for Callaway. Since the zero reference for the narrow range level instrumentation is nominally 96 inches above the top of the highest tube, the 4% value specified fcr MODES 3 and 4 was chosen since it is used throughout the EOPs for heat sink indication and is familiar to the operators. In the main control room there is one Class 1E wide range level indicator per SG and there are four narrow range level indicators per SG, of which three per SG are Class 1E. The "or equivalent" phrase would allow the use of wide range level instrumentation in MODES 3 and 4 in the unlikely event all narrow range level instrumentation were unavailable for a required SG; in MODE 3 this unEkely scenario would result in ITS LCO 3.3.3 non-compliance and would invoke Required Action (s) under PAM Instrumentation. Conversely, the "or equivalent" phrase would allow the use of narrow range level instrumentation in MODE 5 if the one wide range level indicator per SG were unavailable. This flexibility is similar to the approach under which Vogtle was licensed wherein their MODES 3-5 RCS specifications required SG water level to be above the highest point of the SG U-tubes. We are specifying water levels that ensure the same, yet allow the use of all available instrumentation. Before the "or equivalent" instrumentation were used in a given MODE, process measurement effects on the alternate
$ instrum:nt's calibrat::d span would be considsred. Dua to tha unlikely l event of e:ther scenario presenting operationallimitations, given the reduced RCS loop requirements in MODES 3-5 and the instrumentation redundancy, we do not see the need for a pre-determined correlation between the wide range and narrow range level indications; however, we reserve the right to exercise that option should the need arise. 4 Wolf Creek reviewed this particular comment for applicability to Wolf Creek and concurs with the use of the phrase "or equivalent" in the ITS and ITS Bases. Wolf Creek believes that it is appropriate to change their , plant-specific value to 6% narrow range (including uncertainties) since it is used throughout the Emergency Operating Procedures (EMGs), it has operator awareness because of the EMG familiarity, and ensures an SG water level approximately 100 inches above the top of the highest SG e tube. Wolf Creek has done a review of the drawings and design documents and has determined that for MODE 5 conditions (th' .Je range instrumentation is calibrated for cold conditions),66% wit. .ange level corresponds to the top of the highest tube, with margins adaed in for instrument loop errors and readability. The need for flexibility to use either narrow range or wide range indication is most evident when placing the SGs in wet layup conditions. The narrow range instruments are "jumpered" to indicate a constant 50% level. This precludes a feedwater isolation signal at approximately 78%. The operators use SG wide range indication to maintain and monitor SG level. Additionafly, the narrow range instruments are calibrated for normal operating pressure and temperature conditions while the wide range instruments are calibrated for shutdown conditions. The Callaway and Wolf Creek ITS Bases have been modified to explain the "or equivalent" phrase. ATTACHED PAGES: Encl. 2 4-2, 4-4, 4-5 Encl. 3A 3 Encl. 5A 3.4-11,3.4-13,3.4-15,3.4-17 Encl. SB B 3.4-26, B 3.4-31, B 3.4-32, B 3.4-33, B 3.4-35, B 3.4-36 Encl. 6A 8 Encl. 6B 6 l
j REACTOR COOLANT SYSTEM ,- HOT STANDBY , i LIMITING CONDITION FOR OPERATION 3.4.1.2 At least :_: 05: Meactor coolant loops listed below shall be 1-14-LS-22
- OPERABLE and at least two of these reactor coolant looos shall be in operation
} 'when the Rod Control System is capable of rod withdrawal and at least one 1 3 reactor coolant loon in onoration when the Rod Control System is not capable) l Lof rod withdrawalft"J !
- a. Reactor Coolant Loop A :nd M :---- : f r'- m ;:n:._c :nd 1-01-LG -
- - - - - ::-2:nt; r;,
- b. Reactor Coolant Loop S :nd !' -- :- :: :': m ;;n::_ _ _ nd
} c. Reactor Coolarit Loh Cjnd ite :::: '": ' :'::- ;;n:::::::nd , d. Reactor Coolan't Loip'D':rd M: :::r - - d ':: ;:n: 1:: end
- j. -- "- :: _ __ f. ; r .
, APPLICABILITY: MODE 3." 4 44-M Lx 1 ACTION: ! } a. With @equired reactor coolant loops-QPERAIN.E ' 1-19-M ' t linoperspie3 restore the required loops to OPERABLE status within 72 hours or be 1 ? in HOT SHUTDOWN within the next 12 hours. A
- b. With only one reactor coolant loop in operation,4estese eNeest 1 14-LS-22:
'^^^ ':^ : ': 1:' ' ^^ ^ 72 h: - :' .??.1-
- h::: : :- 'h d ' ' - - -
-'t L. 5:- '---4nd the rod control system capable of f rod withdrawal, within 1 hour restore two loops to operation or plac]e .
g the rod control system in a condition incapable of rod withdrawal ;
~
- _ c. Withifour RCS loops inoperable olno reactor coolant loop in operation. 04-IE-29' l immediately place the rod control system in a condition incapable of rod) '
, [withdrawalJIsuspend t all operations involving a reduction in boron concentration 1-19-M i of i_ the Reactor Coolant System and M:' ' '; initiate corrective action to retum the
- Feepuwed one reactor coolant loop to OPERABLE status and operation.
i SURVEILLANCE REQUIREMENTS 4
; 4.4.1.2.1 At least the above required reactor coolant pumps, if not in l operation, shall be determined OPERABLE once per 7 days by verifying correct
{ breaker alionments end indicated power availability.
- 4,or equivsle,#) } Q3A&2,43.45-3 \
1 e.e.l.z.z Tne required steam generators shall be determined OPERABLE by i v ' ing secondary side wede-[narrowirange water level to be greater inan or equal to 1-15-M n i at least once per 12 hours. 4.4.1.2.3 ^' ':::' ' - [The required] reactor coolant loops shall be verified in operation 1-14-LS-22 ' g g_. . ^' ; :--" : :'r.::t least once per 12 hours. 1-01-LG '
- 'All reactor coolant oumps rray be_d
- :n- ;':: removed from operation]
'1 16-Al l for up to 1 hour [per 8 hour perloct)tovided: (1) no operations are permitted that would cause ainution of the Reactor Coolant System boron concentration, M*
and (2) core outlet temperature is maintained at least 10*F below saturation temperature. e__e__ _.,__,,____.m_ e__m_..__ ,, . n 2 104-M3 "No RCP shall IEstarted with any RCS cold leg temperature 5368'F unless the l 1-05-M " secondary side water temperature of each steam generator is s 50*F above each - [ of the RCS cold log temperatures, j WOLF CREEK - UNIT 1 3/4 4-2 Mark-up of CTS 3M.4 5/15/97
l l l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS l l l 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pump, if not in operation, shall be deterrnined OPERABLE once per 7 days by venfying correct breaker alignments and indicated power availability. l l 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by v ing secondary side WncrrowJange water level to be greater than or equal 1-15-M to at least once per 12 hours. l i 4.4.13.3 At least one reactor coolant or RHR loop shall be verified in l operatior. :nd :!:-~' '-.;:::1: :::' -.; at least once per 12 hours. 1-01-LG i 1 { Q 3.4.5-2.1 k squAed l o s.4s .33 i l t i l 1 WOLF CREEK - UNIT 1 3/444 f Mark-up of CTS 3M.4 5/15M7 1
l ! REACTOR COOL. ANT SYSTEM COLD SHUTDOWN - LOOPS Fit i pn , LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At le t one residual heat removal (RHR) loop shall be OPERABLE and i in operation and either: T148:LS4f' l L Wd%skahd/JM
- a. One additional RHR loop shall be OPERABLE #, or .
I
?.6*Ep; or km*V alen} Q 3.4.6-2.1 4
- b. The secondary side w el of at le_ ener 434#83 1 ,
shall be greater than of tt:'
% "%.nge.
J145 W J f 2::maaman i APPLICABILITY: MODE 5 with reactor coolant loops filled ##. L L 6GTjQN: l ) l a. Vvitn one of the RHR loops inoperable and with less than the required i l steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.
- b. With[ required RHR loops inocerable oBno RHR loop in operation, 7 1;10 M C .
suspena en operations involving a reducten in boron concentration of ~ ~ ~ ' ' i j
- the Reactor Coolant System and ;,,..- ' ; initiate corrective action to retum the required RHR loop tc[ OPERABLE status and)peration SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators
. when required shall be determined to be within limits at least once per 12 hours.
~
4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation 4ad 7101-LG"~
""""~d
_ " " .;; :--- - i .: :t least once per 12 hours. . (NEW) Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation at least once per 7 days. [1}11-Mi
#One RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
##A reactor coolant pump shall not be started [with any RCS cold leg temperature] f1-05-M*!
' ' ' * * ~ " "
' unless the secondary water temperature of each steam generator is less n O'F above each of the Reactor Coolant System cold leg temperatures.
*The RHR Dumo may be 2:n.,, - fremoved from ooerotion)for up to 21-16 AT
.. 1 hour [per 8 hour period]provided: (1) no operatons are permitted that would cause %g' dilution of me Reactor Coolant System boron concentration, and (2) core outlet &a. _ 4. -
temperature is maintained at least 10*F below saturation temperature. {** All RHR loops may be removed from operation during planned heatup to MODE 4 106 LS-21 [ when at least one RCS loop is in operation. a l WOLF CREEK- UNIT 1 3/44-5 Mark-up ofCTS3M.4 SA587
l CHANGE ' NUMBER 82iG DESCRIPTION I l 1 11 M This change adds a new surveillance for verification of breaker alignment and power availability to the required pump not in~ operation. This change is in conformance with NUREG 1431 Rev. 1. l 1 12 M The Actions are changed to separate the required actions l for only one required R9R loop OPERABLE and no required l RHR loops OPERABLE. These revised Actions are consistent with the Actions which are required under this LCO in NUREG-1431 Rev. 1. and are more conservatte than current required actions. 1 13 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). l- 1 14 LS 22 The L and ion are r sed ' req 're o yt 17 I OP LE th t loops - oper ion nt ro con ol yst s ca le of od wit rawal o 1 in o ation en th rod co 01 sy em i not pa eo i od wit rawal This c nge i onsi ent th Rev. . I NSEAT 3A -3 E 1431) , 1-15 M A steam generator (SG) level corresponding to 10% of the wide range does not cover all of the SG tubes. To qualify as a valid heat sink, the tubes must be covered. This is a more restrictive change. [,] . 1 16 A Consistent with the intent of traveler TSTF 153, this change revises the note that permits up to 1 hour "deenergization" of RCP/RHR pumps. The revised wording clarifies the intent of the note to allow the pumps to be
" removed from operation" instead of "deenergized", thus permitting other means of removing the pumps from service.
- With this change the pumps are not reauired to be deenergized to use the note (e.g. the pumps may be l ,
isolated, etc.). The change is considered to be administrative because from the standpoint of providing an exception to the LCO requirements (to maintain the operability and operation of the pumps), the revised wording is equivalent. 1 17 LG Not applicable to WCGS. See conversion Comparison Table (Enclosure 3B). WCGS-Description of Changes to CTS 3M.4 3 S/15M7 L
- . . _ _ _ . _ _ _ ,__m. __ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ . . ___ INSERT 3A-3a , 0 3.4.5-1 The LC0 and ACTION b of Specification 3.4.1.2 " Reactor Coolant System, Hot Standby " would be revised to require that two reactor coolant loops be OPE RAB LE. Loop operation requirements would also be revised to be conti,igent on Rod Control System status. The requirement to have a third OPERABLE reactor coolant loop would be deleted, consistent with NUREG-1431, This is acceptable because the MODE 3 decay heat removal requirements are sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay heat. A second RCS loop ensures redundant capability for decay heat removal. When the Rod Control System is capable of rod withdrawal, two loops must be in operation to ensure accident analysis a'sumptions are satisfied. When rod withdrawal is precluded, only one loop is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident analyses which assume only two RCS loops in operation include the Uncontrolled RCCA Bank Withdrawal from Subcritical and the hot zero power RCCA ejection events. The initial conditions and analysis assumptions for these events will be unchanged since two loops must still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal. These reactivity transients rely on the Nuclear Instrumentation System's high flux trips for event termination which occurs very rapidly (on the order of seconds). There would be no benefit of having a third RCS loop OPERABLE for these transients since by the time the loop could be brought into operation, the event would be over for all l practical purposes. INSERT 3A-3b 0 3.4.5-2 , 0 3.4.5-3 Six percent of the narrow range span is specified at the higher temperatures l of MODES 3 and 4 whereas 66% of the wide range span is specified for MODE 5. Both values ensure SG tubes are covered. The Emergency Operatir g Procedures ; cite the 6% narrow range level to ensure heat sink adequacy. l l l l l l l l
.. .-. . ._ . . . . - . ~ . - . _ . - - . - . - -.. ... _ --. . . . . - . . - - - -
l l RCS Loops-MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.4.5.1 Verify required RCS loops are in operation. 12 hours 65 nn SR 3.4.5.2 Verify steam generator secondary side water 12 hours levels are a W4N'for required
, g' RCS loors. T- 3.4-k o{ct"* mI'"') 63.4,5-2.1 43 4.5 3 l SR 3.4.5.3 Verify correct breaker alignment and indicated 7 days l power are available to the required pump that l 1s not in operation, i
l l WCGS-Mark-sq ofNUREG-1431 -ITS 3.4 3.4-11 5/158 7 \ l _ _ m - - - - - -
I RCS Loops MODE 4 l 3.4.6
)
ACTIONS (continued) C0lOITION REQUIRED ACTION COMPLETION TIME I Br 0.; r; wired "l% lc,ep ir.:,pril;. B-1 ";
, ir, = C 5. 24 bar: E ele Tw r; gir;d "00 l=p i;.;,Ritl;.
GB. Required RGG-er-fMR GB.1 Suspend all operations Immediately gjggg loops inoperable. involving a reduction of RCS boron l E concentration. j No RCS or RHR loop in alE operation. 1 GB.2 Initiate action to Inmediately restore one loop to OPERABLE status and operation. I J SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. 12 hours ney r@
.SR 3.4.6.2 Veri SG secondary side water levels are z W 12 hours ggy for required RCS loops. 3,*,M Q 345 -2.
@= of MWY8_ _ Q.3.4.5=3 (continued)
WCGS-Mark-sp ofNUREG-1431 -ITS3.4 3.4 13 5/1587
l l l RCS Loops HN ES. Loops Filled i 3.4.7 1 1 i l 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-H0DE 5 Loops Filled 1 l I LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in I operation, and either: Q3A.5 2.'
- a. One additional RHR loo shall be OPERABLE: or 43.45-3 b.
dde fM The secondary sid water level or at least two steam 1 AMBh~sd generators (SGs) shall be a W __ Mpew
- 1 66% or -e- - - -^ - - - q mu t ea+)
. . . . . . . . . . . . . . . . . . . . . . . . . . N0TES ---- - ----- ---
M) !
- 1. The RHR pump of the loop in operation may be & cr.crgi;cd $314WM removed lfromloperationffor s 1 hour per 8 hour period provided:
- a. No operations are permitted that would cause reduction i of the RCS boron concentration; and
- b. Core outlet temperature is maintained at least 10*F I l below saturation temperature.
l l 2. One required RHR loop may be inoperable for up to 2 hours I for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
- 3. No reactor coolant pump shall be started with ene-ee-mere psiIdhan any,RCS cold leg temperatures s-N52F368'F unless the b(BsPSIP secondary side water temperature of each SG is s 50*F k 18b W, above each of the RCS cold leg temperatures.
- 4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.
l APPLICABILITY: H00E 5 with RCS loops filled. 1 l l l WCGS-Mark-up ofNUREG-1431 -ITS3.4 ~ 3.4 15 5/158 7 l
RCS Loops HODES, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY j SR 3.4.7.2 bkbb Verify SG secondary side water level is :t-47 h[iI-Ii"^ I 12 hours
@ktlarfWwc3:ange)in required SGs. EM M ~ E c.(aWalev % ) 3 4 SR 3.4.7.3 Verify correct breaker aiignment and indicated 7 days power are available to the required RHR plap that is not in operation. -
l l l WCGS-Mark-up ofNUREG-1431 -ITS 3.4 3.4 17 5/158 7
. ~ - . .. . . _ _ . . . - . . . -. -. .. ._ ---
l RCS Loops - H0DE 3 8 3.4.5 l BASES SURVEILLANCE SR 3.4.5.1 (continued) REQUIREMENTS considering other indications and alarms available to the operator in
- the control room to monitor RCS loop performance. !
SR 3.4.5.2 l l SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is veri f ' c,yg or by ensuring that the secondary side narrow range water level is a for required RCS loops. If the SG secondary side narrow range 9uivaled water level is <M the tubes m y become uncovered and the associated 1 p may not be capable of provid ng the heat sink for removal of the l decay heat. The 12 hour Frequenc is considered adequate in view of
\Q3 os.46-3\ A.5-2.)06 */.
other indications available in the control room to alert the operator to a loss of SG level . mThe scle.rme tevelindrumenhMn mtg) i
- u. sed M MODE 5 3 m 4 ib the event all naw l SR 3.4.5'3 F#"6# 'w.iramenta: tion weve umavidM4 (fi: r a. r%uiceA SCr.
Verification that the required RCPs are OPERABLE ensures that safety l analyses liuits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required RCPs. REFERENCES Nene- 1. USAR, Section 15.4.6
- 2. NRC letter (W.' Reckley to N. Carns) dated November 22, 1993: " Wolf Creek Generating Station - Positive Reactivity Addition: Technical Specification Bases Change."
l l l l WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 26 S/1/S/97 i
RCS Loops-H0DE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.2' 64, or equ[ vale"h 3 REQUIREMENTS -- 6% #f (continued)- SR 3.4.6.2 requires verifica ion of SG OPERABILITY. SG OPERABILITY is verified by en 'ng that the seco ry side narrow range water level is 2 .orrequiredi loops. If the SG secondary side narrow range water level is < the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour Frequency is considered adequate in view of other indications available in the f.cntrol room to alert the operator ' to the loss of SG level. - twe. Mas. r law.l inshumentab masEuS*L SR 3.4.6.3 in Meous a + A*w.e.vsnt, att nm W MmocnM.*H went wnavailake. fora ruErud
~
Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES None- l'. USAR,~Section 15.4.6
- 2. NRC letter (W. Reckley to N. -Carns) dated November. 22, 1993: " Wolf Creek Generating l Station _ Positive Reactivity Addition:~ Technical, Specification'_ Bases' Change."
'WCGS-Mark-up ofNUREG-1431 - Bases 3.4 8 3.4 31 S/1/Si97
, RCS Loops-MODE 5 Loops Filled
- . B 3.4.7 l
B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.7 RCS Loops-MODE 5. Loops Filled l BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the ' reactor coolant is the removal of decay heat and transfer of;this heat either to the steam generator (SG) secondary side coolant ! via'nituralTc[rctilationi(Ref. 3) or the component cooling water via the residual heat removal (RM) heat exchangers. While the i principal means for decay heat removal is via the RE System, the SGs are specified as a backup means for redundancy. Even though i the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained voltme of 4 secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, bo;ic acid. In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RR loops connected to the RCS, each loop containing an R m heat exchanger, an R m pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One RE pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification ibut .is. not' sufficient for_.the boron dllution analysis: discuss.ed below. . The number of loops in operation can' vary to suit the operationsl needs. The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal. The LCO provides for redundant paths of decay heat removal capability. The first path can be an RHR loop that must be OPERABLE and in operation.' The second path can be another OPERABLE RHR loop o maintaining two SGs with secondary side water levels above to provide an alternate method for decay heat removal via~_._ uralicittulationJ(Reft 3). Q 3.4. 7 2. I N* Y'f U Q 3. 4.5-3, (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.4 B 3.4 32 5/1/5/97
.. - - - - . - . -- . . - . _ . . - - . - ~ - . - - . _-. . ~ . - . = .
1 INSERT 3A-7a 0 5.2-1 02-22 A The Radioac' ave Effluents Controls Program is revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3 are applicable to these activities. These statements of applicability clarify the allowance for ! surveillance frequency extensions and allowance to perform I missed surveillances. Generic Letter 89-01, " Implementation of Programmatic Controls for Radiological. Effluent Technical Specifications and the Relocation of Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents Control Program in the Administrative Controls Section of the Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 89-
- 01. This change is considered an administrative change since the changes are in the presentation method only. This change is consistent with NUREG-1431 as modified by TSTF 258.
INSERT 3A-7b WC 5.0-004 02-21 A Amendment No.106 for Wolf Creek incorporated a footnote to allow ; the volumetric and surface examination of the RCP "D". motor . l flywheel for the first 10-year ISI interval be delayed for one i operating cycle. The examinations are completed during the ninth l refueling outage. Since the footnote is a one-time exception and has been satisfied, the footnote is no longer applicable and can l be deleted. I i I l
Page 5 of 8 CONVERSION COMPARISON TABLE - CURRENT TS 6.0 APPLICABILITY TECH SPEC CHANGE C0HANCHE PEAK WOLF CREEK CALLAWAY DIABLO CANYON - r NLHBER DESCRIPTION No. Not in CTS. Yes Yes Change the Diesel Fuel Oil Testing Program description for No. Not in CTS. 02-16 A sampled properties of new fuel oil from "within limits" to
" analyzed" within 30 days following sagling and addition of the fuel oil to storage tanks. This wording more clearly defines that within 30 days following the initial new fuel oil sample. the fuel oil is analyzed to establish that the other properties specified in table 1 of ASTM D975-81 are met. This change is consistent with the Bases for ITS SR 3.8.3 _zb[q g.g.: J No. See Section No. LAR submitted Yes nt Pug Flywheel" is being revised consistent Yes 02-17 " Reactor C 3/4.4. CN 10-03-LS 12/3/96.
LS-1 with The proposed changes provide an exception to the examination requiremer.ts in Regulatory Guide 1.14. Revison 1. " Reactor Coolant Pug Flyidal Integrity." Revise the Radioactive Effluent Controls Program dose rate y blunffy f(CPIr. Yes Yes(QG2-1} 02-18 hts A limits to reflect changes to 10_CFR Part 20.)i dra -~_ '//es hnytf Lejtet ajyM prped %1er.f j_ [Qs.2.-I{ Yes Yes The surveillance interval for verifying that other No. Addressed in No. Addressed in 02-19 3/4.8 (See CN 01-properties are within limits for ASTM 20 fuel oil is 3/4.8 (See CN 01-LS-2 60-LS24). changed from "within 30 days" to "within 31 days" after 60 LS24). obtaining a sagle. Yes Yes No. Not in CTS. No. Not in CTS. 42-20 Add the provisions of Specification SR 3.0.2 and SR 3.0.3 A are applicable to the Diesel Fuel Oil Testing program. This change is consistent with TSTF-118. Yes Yes Yes Yes 03-01 The method for submitting all reports is revised to be in A accordance with 10 CFR 50.4. 0 1-21 14wAT 32> -5bp Mc 64-N+ ) fo2-21 adssidr3e,- sa. O s.2-1 l LA % o2-23 msE*: T ss- Sc-Mtc. s.o-co4l S/1SM 7 WCGS-Camversion Comparison Table- CTS 3M.0 t
INSERT 38-Sa 0 5.2-1 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-22 The Radioactive Effluents Controls Program Yes Yes Yes Yes A is revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3 are applicable to these activities. INSERT 3B-5b WC 5.0-004 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-21 Amendment No.106 for Wolf Creek No No Yes No A incorporated a footnote to allow the volumetric and surface examination of the RCP "D" motor flywheel for the first 10-year ISI interval be delayed for one operating cycle. The examinations are completed during the ninth refueling outage. Since the footnote is a one-time exception and has been satisfied, the footnote is no longer applicable and can be deleted. INSERT 3B-5c DC 5.0-004 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMMANCHE PEAK WOLF CREEK CALLAWAY 02-23 DCPP Administrative Programs, CTS 6.8.4.d. Yes No No No LG " Backup Method for Determining Subcooling Margin." and 6.8.4.f, " Containment Poar and Turbine Building Cranes." were evaluated for reloaction outside the TS to a licensee-controlled document consiste with 10 CFR 50.36 screening criteria.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 5.0-005 APPLICABILITY: WC REQUEST: NRC letter dated February 27,1998 reissued CTS page 6-18b due to an administrative error which inadvertently omitted a sentence from CTS 6.8.5. The CTS has been marked up to reflect the issuance of page 6-18b. ATTACHED PAGES: Encl. 2 6-18b l I i I 4
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) The maximum allowable containment leakage rate, L., at P., shall be 0.20% of containment air weight per day. Leakage rate acceptance enteria are:
- a. Containment leakage rate acceptance enterion is s1.0 L.. During the first unit startup following testing in accordance with this program, the leakage rate acceptance entena are 50.60 L. for the Type B and C tests and 50.75 L. for Type a tests;
- b. Air lock testing acceptance enteria are:
- 1) Overall air lock leakage rate is s0.05 L. when tested at 2P.;
- 2) For each door, leakage rate is 50.005 L. when pressurized to 210 psig.
The provisions of Technical Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The provisions of Technical Speerfication 4.0.3 are applicable to the Containment Leakage Rate Testing Program. 7 6.8.5 a. Explosive Gas and Storace Tank Radioactivity Monitorina Procram This program provides controls for potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks The program shallinclude: f
- 1. The limits for concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM and a surveillance program to ensure the limits are maintained.
- 2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 20.5 rem to any individual in an UNRESTRICTED AREA in the event of an uncontrolled release of the tanks' contents, consistent with Branch Technical Position ETSB 11-5," Postulated Radioactive Releases due to Waste Gas System Leak or Failure."
(o.8.5 "Tha followin3 pegrams , reloc.ated from b.a. ~Tedvvu.c:A. l '# b Spa.cVficMirns, b uSAR cha@ %, sbit ba. iU p ie-e fed. x A na k+ > k cA. : _ l l l WOLF CREEK UNIT 1 6-18b Amendment No. M 101 Mark-up of CTS 6.0 S/15/97 l I
l l ADDITIONAL INFORMATION COVER SHEET I ADDITIONAL INFORMATION NO: WC 5.0-006 APPLICABILITY: WC REQUEST: ITS Section 5.5.11 is revised to correctly reflect that the Control Room Emergency Ventilation System - Filtration is part of the ITS 5.5.11b. program. Also editorial changes are made for I_ consistency. l; l ATTACHED PAGES: Encl. 5A 5.0-22, 5.0-23 l I i 4 I f
Programs and Manuals 5.5 5.5 Programs and Manuals ! l 5.5.11 Ventilation Filter Testina Procram (VFTP) (continued) i ESF Ventilation System Flowrate i Control Room: Emergency . Ventilation System Filtration 200.0. .cfm EMD.. Control RUEmergency Ventilation System _ , Pressurization ' 750'cfm l Auxillary /FUerBuiTding Emergency Tihaust~ ~6500'cfm
- b. Damonstrate for each of the ESF systems that an inplace test of the charcoal absorber shows a penetration and system bypass
< [0.05]% when tested in accordance with Regulatory Guide l'.52', pgg Revision 2. ;r.d AS". "510 1000 at the system flowrate specified below E* 10*3
-_ ESF Ventilation System Flowrate_
(Chhol Room Emergene Ventjihn @ Tea - F;ltvAim 2ccocb b {"'#****6I Control /no cfm
- MlaMRoom Emergency iigi@Ventilation Edimist System - Pressurization
~~ ~ ~
@[shi
- c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2. shows the methyl iodide ggg penetration less than the value specified below when tested in accordance with ASTM D3803 1989 atatemperatureof@30*Cand ~
gggpsg , greater t.:an or equal to the relative humidity specified belowl.
%c s.o-eoj ESF Ventilation Sy hs Penetration _ _ _ RH
(( FiwWm /1%mivinkim)J l uc. s.o-ook l Control Room Emergency Ventilation System 4 22 70% g g:psy , Auxiliary / Fuel Building Emergency Exhaust 2% 70%
' R;n s;c'; Netc. Alic;21; pretr;;ica - [100t crathyl iodid; cfficicray for ch;rs;l c;dit;d in ;teff ;;f;ty cv;1;;ti;n]/ (;;f;ty fater). 0;f;ty f;;ter:5] for ;y;t;;.; with hata;.
- U] f;r ;y;t a.; with;;t h;;tcr;.
(continued) WCGS-Mark-up ofNUREG-1431-1TS 5.0 5.0 22 S/158 7
Programs and Manuals 5.5 5.5 Programs and Manuals
- d. Demonstrate atgeast_once per 18 months for each of the ESF !M;5M19 systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal absorbers is less than the value specified below when tested in accordance with Regulatory GQide3~52 JR~efision-2 and ASI N510-1000) at the system 43 BiPS c,g flowrate specified below E* 10%-}.
ESF Ventilation System Delta P Flowrate
@eq4=cy Ve=EWm y.tw ry { uc s.o -ooe.1 Control Room iltration M 6.6 inz W. G. 2000?cfm : !!B4PSs
~'
ControT RoonhPressurization' 3.T in'.. 'W. G'. 750'cfm AuxiTiary/Fual BuiTding Emer@gency Exhaust 4.7'in. W. G .
~65001cfm Q Q % Fbb'wset* M* u c. s .o -co sj r
- e. Demonstrate at least once per 18 months thu the heaters for .25?Ss110 each of3he ESF systems dissipate the value specified below when ~
tested in accordence with ASME N5101975. #85PS5 ESF Ventilation System _ _ _ Wattage EmeFqgvsehbm von bnn Control Room 5 *11 kW ' iBsPS? f Auxiliary / Fuel Building , 37 i 3 kW herp$ Shm] I The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test l frequencies.
+. NsenT s.o -2ry-1 o s .s - 7 )
5.5.12 Exolosive Gas and Storace Tank Radioactivity Monitorina Proaram This program provides controls for potentially explosive gas mixtures tBfPSI contained in the Waste Gas Holdup System, the quantity of g gs, radioactivity contained in gas storage tanks cr fed ;nts the effges . treatment system, and the quantity of radioactivity contained in LBsPST
^
unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities ' hall be determined following the methodology in Branch RB PS Technical Position (BTP) ETSS 11-5, Revision 0, July 1981,
" Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Revision 2, July 1981 Section 15.7.3, " Postulated Radioactive Release due to Tank Failures."
(continued) WCGS-Afark-up ofNUREG-101-ITS S.O 5.0 23 S/1S/97}}