ML20236Y277
| ML20236Y277 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/05/1998 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20236Y268 | List: |
| References | |
| NUDOCS 9808120040 | |
| Download: ML20236Y277 (400) | |
Text
{{#Wiki_filter:SDM-T,,, r 200*r B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)-T.,,-+-2994
-BASES BACKGROUlO According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel, SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (A00s).
As such, the SDM defines the degree of suberiticality that would be obtained immediately following the insertion ee-seren of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn. The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core suberitical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor _f Coolant System (RCS). The' M ]Gg g g System can compensate for Ct io'i ~~ the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to L no load. In addition, the,RijMReg System, together with the boration system, provides the SDM during power operation and is
@ 3'*I capable of making the core suberitical rapidly enough to prevent exceeding acceptable fuel damage limits, assiming that the rod of highest reactivity worth remains fully withdrawn. The W .
eBNppluglggttel'gsttui 2111: :;;,;c,.. ;;y= can m I sgit[e_~"tgili$n"concentrationZto compensate for fuel depletion during operation and all xenon burnout reactivity changes and can maintain the reactor subcritical under cold conditions. During power operation, SDN control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.76, " Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. (continued) WCGS-Mark-up ofNUREG-1431-Bees 3.1 B 3.1-1 $/1587 9908120040 990905 ? PDR ADOCK 05000492 ! P PM j
l NTC B 3.1.43 B 3.1 REACTIVITY CONTROL SYSTEMS L B 3.1.43- Moderator Temperature Coefficient (MTC) BASES BACKGROUlm According to GDC 11 (Ref.1), the reactor core and its ! interaction with the Reactor Coolant System (RCS) must be l designed for inherently stable power operation, even in the !- possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.
'The NTC relates a change in core reactivity to a change in reactor coolant temperature (a positive NTC means that reactivity r
increases with increasing moderator temperature; conversely, a 1 negative NTC means that reactivity decreases with increasing moderator temperature). "; renter i; i;ig d M eg.r;t .;ith
; =;;;;in .2 :=r ;M hiF;; i;n;ib1; rar.,. ;f f=1 ;;;1e l l ;;;r.ti;r.. Therefore,;~9ftIFEM a coolant temperature l l
increase will cause a reactivity decrease, so that the coolant I temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result. i- l MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by i measurements. Beth-4M44el-end Reload cores are designed so that ! l the beginning of cycle (BOC) HTC is less than zero when THERMAL POWER is at RTP. - The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an NTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) NTC is also limited by the requirements of the accident analysis; ami fFuel cycles tMt ;.r; i;;F.;d te shi;; hi;;h h .. ;;; ;r tMt b; ;L..F; te ,/Rr ;Mrat;ri; tis are evaluated to ensure that the NTC does not exceed the EOC limit. The limitations on NTC are provided to ensure that the value of this ficient remains within the limiting conditions assumed l in t accident and transient analyses, i Q 3.t.G:-l l redh4c." u" (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 14 S/15m
HTC B 3.1.43 BASES BACKGROUND If the LCO limits are not met, the unit response during (continued) transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity. The SRs for measurement of the MTC at the beginning and near the i end of the fuel cycle are adequate to confirm that the NTC l remains within its limits, since this coefficient changes slowly, l due principally to the reduction in.RCS boron concentration associated with fuel burnup. APPLICABLE The acceptance criteria for the specified MTC are: SAFETY ANALY5ES
- a. The MTC values must remain within the bounds of those used l in the accident analysis (Ref. 2); and
- b. The NTC must be such that inherently stable power I operations result during normal operation and accidents, gg such as overheating and overcooling events.
. u. 4 G 5 1-(s-83 T , Chapter 15 (Ref. 2), contains analyses of accidents tha result in both overheating and overcooling of the reactor core. NTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).
The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 2 4) or RTP, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature. In order to ensure a bounding accident analysis, the HTC is assumed to be its most limiting value for the analysis conditions (continued) WCGS-Mark-up ofNUREG-lui-Bases 3.1 B 3.1 15 S/15n7
MTC B 3.1 43 BASES SURVEILLANCE SR 3.1.43.2 xd f? 2.1.4.2 (continued) REQUIREMENTS l'. The_SRJ s_notirequi red ' to_bepformedionceisachigjcl_e
*tthirPTeffectivenfu11 'oewerDdays7EFPDsEafter" reaching t!nQi[Rivalint'ofian.Jguf11btfun'RTPlalDuds 3ist3ARD)
_.rconcentrat piot3co m 3.l . Gt = l 4 2; If the 300 ppm Surveillance limit is exceeded it is possible that the EOC limit on MTC could be reached before the planned EOC. Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit. Q.11.6r l } Jr 3. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is less negative er; paiti n than the 60 ppe Surveillance limit, the E0C limit will not be exceeded because of the gradual manner in which MTC changes with core burnup. REFERENCES 1. 10 CFR 50 Appendix A GDC 11.
- 2. FfiAR USAR, Chapter f15}.
- 3. '.0^." 0270 "" ", ~.t.;tirg.;,a; 'ulnd 0;fet3 En12 tim
- t:L,dr,legy," Aly 1005. IISAG!OO77' D ]ohd Safet9 l Evaluation. Methodology forithe WolfiCreek _ Generating Station."_
- 4. i = , 0.'.c.pter [153 1
i l l l l l I WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1-20 S/158 7
Control Bank Insertion Limits B 3.1.76 BASES [7 ).'Og 5 -- ~.-^.F^-h1 E "N. LA 0,d = 22-2.
. #,. '; h BACKGROUND 6*e.wg. ..yT;itMrd 7.t1 p;ition& q.07 . - g.r. .;.,
of = m p.
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(continued) p;ition i; d; fir d in tM C0' Jt. TM Sily wit'.dr;w9 .1,33 l*-'l The control banks are used for precise reactivity control of the reactor. The positions of the control banks gpn2 er; ei. ;11y controlled MJgDEig' automatically by the Rod Control System. Lt ;;n ;1 n k n r.211y ar.trell;d. They are capable of adding reactivity very quickly (compared to borating or diluting). De M1 tights ~allst2eimaint21as~ d ahne~(Iss2BMm34gts MMineerrtpiie"11STiv63lremMtWMR$nglimprupi] WPasierEbreittbri The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.54, "EntrEgyN!05mflis3 LCO 3.1.65, " Shutdown Bank Insertion Limits," LCO 3.1.76, Jpgr;ol'Ep!fBasertion 13 rip 2 Z LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD) " and LCO 3.2.4,
" QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.
The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained. Operation within the subject LCO limits will prevent fuel : cladding failures that would breach the primary fission product i barrier and release fission products to the reactor coolant in I the event of a loss of coolant accident (LOCA), loss of flow, I ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function, l APPLICABLE The shutdown and control bank insertion limits, AFD, and j l SAFETY ANALYSES QPTR LCOs are required to prevent power distributions that could I result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring temination by an RTS trip function. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1-38 S/1S/97
N 231 (17.231) (67,231) ANKB FULLY WITHDRAWN j 200 [ (100,190) 7 (0,191) y 'N
!S.
Z 150 / 9 / 5 2 BAN C 100 ' to BANK D o (0, 73) b 50 THIS FIGURE FOR
\ ILLUSTRATION ONLY.
F LYINSERTED 0 NOT USE FOR O ERATION. l 0 (19.0) , g l ! 0 20 40 60 80 100 PERCENT OF RTP Figure B 3.1.7-1 (page 1 of 1) l C trol Bank Insertion vs. Percent RTP Weted-Ceumpl<. Oyr-) {Q3.t.ce-1 , WOG STS B 3.1-45 Rev 1, 04/07/95
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.1-1 APPLICABILITY: WC, CA REQUEST: 3.1.1 Shutdown Margin (SDM) (Wolf Creek & Callaway) DOC 01-02-M CTS 3/4.1.1 Applicability ITS 3.1.1 Applicability l Comment: According to the Conversion Comparison Table, " MODE 2 with Keff < 1.0" . and " MODE 5" are added to the Applicability section of TS 3.1.1 for Wolf Creek and l Callaway. All of the FLOG ITS Sections 3.1.1 have these applicability requirements included in the ITS and not in the CTS. An inadequate justification for these changes is provided. Provide a discussion explaining / justifying these changes. FLOG RESPONSE: The " MODE 2 with k , < 1.0" Applicability was added under DOC 01-02-M, i applicable only to Callaway and Wolf Creek. The " MODE 5" Applicability ! l was added under DOC 02-01-A, applicable to all FLOG plants. While all FLOG plants have the same ITS 3.1.1 Applicability, the statement that the above MODE requirements are not in the CTS is untrue for Diablo , Canyon and Comanche Peak whose CTS 3.1.1.1 and 3.1.1.2 include MODES 1-5. Those plants revised their CTS SDM LCO Applicability { l based on DOCS 01-06-A and 02-01-A. I l The change being made for Callaway and Wolf Creek under DOC 01 l M ensures that the SDM LCO Applicability covers that portion of MODE 2 not covered by ITS 3.1.6, " Control Bank insertion Limits." Adding " MODE 2 with k,n < 1.0" to the 3.1.1 SDM LCO Applicability covers the period of time that the control banks are withdrawn prior to reactor criticality and entry into ITS 3.1.6. DOC 01-02-M is revised to read as follows:
"The proposed modification redefines the Applicability of the Specification-to include " Mode 2 with k,n < 1.0" in addition to Modes 3,4, and 5 (see CN 02-01-A). The current Specification for control bank insertion limits (ITS LCO 3.1.6) and shutdown bank insertior limits (ITS LCO 3.1.5) defins the Shutdown Margin requirements for Mode 1 and Mode 2 with k,, > 1.0. ,
The SDM Applicabi!ity requirement added to CTS 3.1.1.1 ensures sufficient negative reactivity is available to meet the assumptions of the safety analyses throughout all of MODE 2. The proposed change would be more restrictive, but would represent only a small change from the , current SDM Applicability requirements." j i ATTACHED PAGES: Encl. 3A 1 l 4
DESCRIPTION OF CHANGES TO CURRENT TS Section 3/4.1 This enclosure contains a brief description / justification for each marked-up change to current Technical Specifications. The changes are identified by change numbers contained in Enclosure 2 (Mark up of the current Technical Specifications). In addition, the referenced No Significant Hazards Considerations (NSHCs) are contained in Enclosure 4. Only technical changes are discussed: administrative changes (i.e., format, presentation, and editorial changes) made to conform to the improved Technical Specifications are not discussed. For Enclosures 3A, 3B, 4, 6A and 6B, text in brackets "[ ]" indicates the information is plant specific and is not comon to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUMBER HSt!C DffibPTION TA3.1-oos.} 01 01 LG In accordance with TSTF 9, , this change would move the specified limit for Shut own Margin (SDM) from current TS to the COLR. This change occurs in several specifications including that for SDH and those specifications with ACTIONS that require verifying SDM within limits. SDH is a cycle-specific parameter that is calculated based on an NRC approved methodology. Moving the SDM to the COLR will provide core design and operational flexibility that can be used for improved fuel management. 01 02 M se modific onMines t applicab li i the spe 1catio o inclu Mode ith kg 1.0" n [ add on to es 3 . and 5 ee CN 02- A). he
) rrent cifica n for trol ban nserti limit (an S Spec' cation . 6) defi s the 5 down rgin plicabi y requi ents for de 1 an ode 2 th
- t 1. The pro ed change uld be e res 1cti ut wo d repre only a s 1 change m th curr SDM applicab' lty requir ts. I4%RT 34 1 o LQ3.11 -
01 03 LS-1 The Action Statement would be modified to reflect that the requirement to initiate boration at a specified rate with fluid at a specified boron concentration is generalized to simply require boration. As described in the ITS Bases, the required flow rate and boron concentration should be ! selected depending on plant conditions and available equipment. The ITS Bases allow the operator to use the "best source available for the plbnt conditions. This is an example of maintaining the overall safety requirement in TS but removing procedural details from the TS allowing the plant operator the ability to select the appropriate procedure and equipment for the existing plant condition. WCGS-Description of Changes to CTS 3.1 1 5/15/97
l l INSERT 3A-la 0 3.1-1 The proposed modification redefines the Applicability of the Specification to include " Mode 2 with k,,, < 1.0" in addition to Modes 3, 4, and 5 (see CN 02-01-A). The current Specification for control bank insertion lin.sts (ITS LC0 3.1.6) and shutdown bank insertion limits (ITS LCO 3.1.5) define the Shutdown Margin requirements for Mode 1 and Mode 2 with k ,,11.0. The SDM Applicability requirement added to CTS 3.1.1.1 ensures sufficient negative reactivity is available to meet the assumptions of the safety analyses throughout all of MODE 2. The proposed change would be more restrictive, but would represent only a small change from the current SDM Applicability requirements. l \ l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.1-3 APPLICABILITY: DC, CP, WC, CA REQUEST: 3.1.1 Shutdown Margin (SDM)(All FLOG Plants) DOC 01-10-M CTS SR 4.1.1.1.1 ITS SR 3.1.1.1 Comment: The justi'ication for modifying applicability of SR 3.1.1.1 is inadequate; it only refers to consistency with NUREG-1431. Also, it is not apparent why this change is not applicable to Wolf Creek and Callaway. FLOG RESPONSE: For DCPP and CPSES, DOC 01-10-M is revised to state the following:
"In the ITS format, the SHUTDOWN MARGIN in MODE 1 and MODE 2 with ke21.0 is controlled through compliance with control rod insertion limits. For those modes or conditions in which compliance with control l rod insertion limits is not required, the SHUTDOWN MARGIN is verified in !
the more traditional manner by consideration of such factors as Reactor j Coolant System boron concentration, coolant temperature;. xenon and samarium concentrations, etc. Thus, the applicability of CTS SR l 4.1.1.1.1.e is modified by this change to be applicable to MODE 2 with k on I
< 1.0 as well as the current MODES 3 and 4. This change is more restrictive, in that CTS 4.1.1.1.1.b addresses MODEG 1 and 2 with k ,
21.0, and CTS 4.1.1.1.1.e addresses MODES 3 and 4. MODE 2 with k,n
< 1.0 is not specifically addressed in the CTS. See also revised Change 01-06-A, which provides a broad discussion of how the applicabilities for CTS 3.1.1.1,3.1.1.2,3.1.3.5 and 3.1.3.6 have been revised."
The Wolf Creek and Callaway Technical Specifications were modified by Amendment 89 and 103 respectively, to contain MODE 3,4, and 5 Specifications for " Shutdown Margin" and a separate MODE 1 and 2 Specification for " Core Reactivity." This eliminated the need for individual MODE applications under the Surveillance Section. Wolf Creek and Callaway used DOC 01-02-M to make the MODE 2 with K.,<1.0 change to both the LCO and the SR. This makes DOC 01-10-M not applicable to Wolf Creek and Callaway (see Enclosure 3B). ATTACHED PAGES: i None I L__ - _ - - -
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-13 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.1.4 Rod Group Alignment Limits (Comanche Peak) j DOC 12-07-A l ITS 3.1.4 Bases Comment: The DOC states, for Required Action B.2.6, that "the ITS Bases discuss the accident analysis affected by rod misalignment." The associated Bases do not list the accident analyses that require re-evaluation, similar to that provided by the other Four Loop Group plants. List in the Bases the accident analyses that require re-evaluation. FLOG RESPONSE: The APPLICABLE SAFETY ANALYSES section of ITS 3.1.4 Bases provides an appropriate description of the various manners in which a misaligned rod can affect the safety analyses. The requirement in ITS 3.1.4 REQUIRED ACTION B 2.6 is to evaluate the safety analyses; the affected analyses are described more fully by the APPLICABLE SAFETY ANALYSES than by the list transported from the CTS. In fact, many of the analyses listed (e.g., Decrease in Reactor Coolant inventory in USAR Section 15.6) are not affected by reasonable rod misalignments; whereas some transients that are sensitive to misaligned rods (most of the Power l Distribution and Reactivity Anomaly accidents described in USAR Section 15.4) are not listed. Because of the potential conflicts between the APPLICABLE SAFETY ANALYSES section and the list of CTS Table 3.1-1, it is preferable to not add the list, but refer to the APPLICABLE ! SAFETY ANALYSES section. The ITS Bases ACTIONS B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 are revised to indicate that the accident analysis presented in USAR Chapter 15 that may be adversely affected will be i evaluated to ensure that the analyses results remain valid for the duration of continued operation. Callaway, Wolf Creek and Diablo Canyon have reviewed this Comment and concur with the above discussion. Their ITS Bases have been revised to delete the list of accident analyses that require re-evaluation and refer to USAR Chapter 15.. ATTACHED PAGES: Encl. 58 8 3.1-28, B 3.1-29 l r st
4 Rod Group Alignment Limits B 3.1.54 BASES ACTIONS B.2.2. B.2.3. B.2.4. B.2.5. and B.2.6 (continued) the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. When a rod is known to be misa11gned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Fre~uency q of 12 hours is sufficient to ensure this requirement continues ~ to be met. Verifying that Fo(Z) and Pan are within the required limits ensures that current operation at 75% RTP with a rod misa11gned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate Fa(Z) and Pn. 3 Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.
, h oNowi _cciden . ly jrequirejevalu ~ _,;fo.r contL ___ rati _w ith isali cod:
Rod' ster: _ _ rol fs _JyfIn togC cteristics- _ __ .f.1 Con ;Assf _y,Misa]F ,g: I) c. ss of _ctor_ _ anLf _ . .11:Rup ' Pipes _or rom Cra inilja. . _ipesi __ M uate- Em_etgenc ~ re l _ in9 3 : I
- d. Si e Pod'El .er Cont . embly]H rawarat' 1
- eri
- e. Haj _ Reactor lant Syst ipe Ruptur
[t (Loss'o(Coolant
\
j t_ ( Accident); - p,q @heaccide.stanalyse.,s pre 3ent_d A, uSAR C.haffe>- i5 NcU 4) 4bC 1 o%
.neu be. actveesa.S o.me;te.ca will Es. c.valuatecs -to ensarv 4ha+ h-
\c -t v. w es. , rue a1sarwes re umsulth stt~ re ma.hm_vac= tic 4._ -ror h watibn of
~_- - -
a h6 (continued) WCGSMark upofNUREG-1431-Bases 3.1 B 3.128 5/158 7
Rod Group Alignment Linits B 3.1."4 BASES ACTIONS 2/ B. d 5 d.4. 2.5.MndB.I6 (co in ) N
/@
G .n _..ry ._lant' TP' __ _ re W
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- AW '._..
-m_ w 1 B-+ C3 When Required Actions cannot be completed sithin their Completion Time, the unit must be brought to a MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without ,
challenging the plant systems. 1 C.1.1 ord C.1.2 D:T.1'and F 1 2 More than one ce;; trol rod becoming misaligned from its group i eveenge demand position is not expected, and has the potential to reduce SDN. Therefore SDM must be evaluated. One hour allows the operator adequate time to determine SDN. Restoration of the required SDH, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases for LCO 3.1.1. The required Completion Time of I hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accirient occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored. C-e D~2 If more than one rod is found to be misa11gned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. S+nec =tca tic (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 29 5/1587
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-14 APPLICABILITY: WC REQUEST: CTS 3.1.3.1 Movable Control Assemblies (Wolf Creek) DOC 12-12-LS-13 CTS 3.1.3.1, Action 4 l Comment: The CTS mark-up applies DOC 12-12-LS-13 to CTS 3.1.3.1 Action 4, which is incorrect. Correct the CTS mark-up/ DOC. FLOG RESPONSE: Applying DOC 12-12-LS-13 to CTS 3.1.3.1 Action 4 is correct. In the WCGS Technical Specification 3.1.3.1 Actions, there is a Table specifying the Cause of Inoperability. Item c) in this Table discusses rods inoperable due to a rod control urgent failure alarm or other electical problem in the l rod control system but trippable with ACTION 4 applicable for both One Rod and More Than One Rod inoperable. DOC 12-12-LS-13 justifies deleting reference causes of control rod inoperability due to rod control urgent failure or other electrical problems in the Rod Control System. Applying DOC 12-12-LS-13 to ACTION 4 on TS page 3/41-9 is consistent with applying the DOC to the Cause of Inoperability Table item c). No changes required. ATTACHED PAGES: None 1 l i l {
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-15 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.1.4 Rod Group Alignment Limits CTS 3/4.1.3 Movable Control Assemblies (All FLOG Plants) DOC 12-14-M Comment: The ITS has changed the wording of the TS from "trippability" to
" operability," and references TSTF-107 which is not yet approved (though it is expected to be approved with the OGs next revision of TSTF-107. The result is that the FLOG plantt have inconsistently incorporated generic changes into the Bases (i.e., the Bases paragraphs for B.2.1.1 and B.2.1.2). This change is a less restrictive change in that it precludes LCO 3.0.3 entry for unforeseen inoperabilities. TSTF-107 needs to be discussed / approved at the next TSTF OG/NRC Meeting, and the FLOG will then need to incorporate the resulting generic TS requirements.
FLOG RESPONSE: It is the FLOG's understanding that EXCEL Services Corporation met with the NRC on May 23,1998 to discuss TSTF-107. The result of that meeting has been reported to be agreement to approve TSTF-107 with a minor Bases change. Revision 1 of TSTF-107 has been incorporated into the FLOG submittals. In the ITS, rod operability is addressed in the Bases as trippability within the drop time requirements of ITS SR 3.1.4.3. If not met, Condition A would be entered which requires SDM verification and shutdown to Mode 3 in 6 hours, which then exits the LCO. In the CTS, the action for an untrippable rod is essentially the same as the ITS. No action is provided in the CTS for discovering in Mode 1 or 2 that a rod would not meet insertion time requirements; therefore, CTS LCO 3.0.3 would be entered. LCO 3.0.3 allows one hour to initiate a shutdown and 6 additional hours to reach Mode 3. Because the ITS only allows 6 hours to reach Mode 3 (instead of up to 7 as allowed by LCO 3.0.3), the change from "untrippable" to " inoperable" in CTS 3.1.3.1 is more restrictive. ATTACHED PAGES: Encl. 5A Traveler Status page , Encl. SB B 3.1-21, B 3.1-24, B 3.1-25, B 3.1-26, B 3.1-27 l
i INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS 4 TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS Special Test Exceptions 3.1.10 is retained and renumbered as 3.1.8, consistent with this traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. f* 8 8-86f I TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. l (TSTF-107, bt. D Incorporated 3.1 6 . le 2.s-8 F l TSTF-108, N::in::g . :;d -NA- Not NRC approved nW Rev.1 _ _ incorporate L 3.L- 2.] ____ ._ Ta s.i.oai 1 l TSTF-110 Incorporated 3.1-10 w Rev. 2. N 8#* - W l-* TSTF-136 Incorporated 3.1-9,3.1-15 (p gc y D lTn.s.t- cor.J TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoff date Nii; ..b. -T;rir ;...;d ener# "I
^
l TSTF-142 si -NA-incorper>f4M _ _
- 3. I - 2.2. _ .g J.,;, W Pt.appre ]
( Incorporated 3.1-7 (WOG3YkN _ incorporated -~ 3.1-16 '/ K .. ,J^ D I@ 3 8- as I
-h) 71P. 3.1 - 00(,)
5/15/97
Rod Group Alignment Ltits B 3.1.54 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.54 Rod Group Alignment Liliits BASES l w s. i- s s- l BACKGROUND The OPERABILITYEMRgM;<$1y5 f tne snutdown and control rods is an Initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial asswnption in the safety analysis that directly affects core power distributions and assumptions of available SDH. The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10. " Reactor Design," GDC 26, " Reactivity Control System Redundancy and "retati;n" CipliblTily;" (Ref.1), and 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2). Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core r>peration in design power peaking limits and the core design requirement of a minimus SDN. Limits on control rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensuro that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDHs). Each CRDH moves its RCCA one step (approximately % inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System. l The RCCAs are divided among four control banks and five shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A benk Of RCCAs censists of tw Fw;;s that (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 21 S/15/97 i
I I Rod Group Alignment Limits l B 3.1.54 ' 1 BASES ; APPLICABLE bounds the situation when a rod is misaligned from its group by l SAFETY ANALYSES 12 steps. l (continued) Another type of misalignment occurs if one RCCA fails to insert l upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the l required SDM is met with the maximum worth RCCA also fully l withdrawn (Ref. 35). The Required Actions in this LC0 ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod ! worth are preserved. Continued operation of the reactor with a misa11gned control rod is allowed if the heat flux hot channel factor (Fo(Z)) l and the nuclear enthalpy hot channel factor (Fa"w) are verified to be within their limits in the COLR and the safety l analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod l insertion limits, AFD limits, and quadrant power tilt limits are , not preserved. Therefore, the limits may not preserve the design l peaking factors, and Fa(Z) and F"w 6 must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of Fa(Z) and t Fa"s to the operating limits. Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of the NP,C P;1 icy Stat; ent 1.0;CFR_50;36(c)(2)(11_); . LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The I requirements on OPERABILITY ensure that upon reactor trip, the l assumed reactivity will be available and will be inserted. The leMd' -(grE2IO'MrequirementsMensure that the RCCAs and banks ( maintain the correct power distribution and rod alignment. The rod,0PERABILI.TY., requirement islsatisfied;provided,the]od is trippable and meets.the rod _ drop time, requirements of SR 3.1.4.3, am. seprake. from h. sijnme q o 3,9 pg, r wemem% MLk. (continued) WCGS-Mark-up ofNUREG-H31-Bases 3.1 B 3.1-24 S/158 7 l
i Rod Group Alignment Limits B 3.1.54 l 1 BASES m LC0 Rod 'controlTmalfunctions tt)at resulgi' y a] (continued) rec (1E ntfittt_ cot 13pihyytsE tut %DJgOtagt ttippabilityi~do:not' necessarily _ result])]3id'1])pperab11.1 ty. The requirement to maintain the rod alignment to within plus or minus 12 steps of~t. heir 5 stoup'~$tto C6bAtR'Rprid][pTfjMn is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed. Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LIRs or unacceptable ; SDHs, all of which may constitute initial conditions inconsistent i with the safety analysis. I l APPLICABILITY Th? requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to i affect the safety of the plant. In MODES 3, 4. 5, and 6, the ! alignment limits do not apply because tM e.,tr;l ret er; l bettened-end the reactor is shut down and not producing fission i powerOl@!MgetioD3!DI!LIEllEReerla.3mytitsarten of_Ig(_if)MtthgL].1 pits as in~lEDF,5,1ggG. In the shutdown ' MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the i RCS. See LCO 3.1.1, "SHLHDOWN MARGIN," fS016-4, r 2007." for SDM in MODES 3, 4, and 5 and LCO 3.9.1, " Boron Concentration " ; for boron concentration requirements during refueling. ' l ACTIONS A.1.1 and A.1.2 p ig j When one or more rods are ingperableh[nJPTf$ e\.there is a possibility that the required Sun may De adverseiy affected. ! I Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is (continued) WCGS-Mar l,-ap ofNUREG-1431-Bases 3.1 B 3.1 25 S/15/97
Rod Group Alignment Lizits B 3.1.54 BASES ACTIONS A.1.1 and A.1.2 (continued) adequate for determining SDM and, if necessary, for initiating emergency boration end to restor +nge SDM. In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maxime worth. L2 If the tete +ppeble inqperabTe rod (s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is ramnable. based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. ; 1 U fos.i-irl When a rod becomesimisa11oned, jt can usually be moved and is still trippable(MM. If the rod can be realigned within the Completion Time oT 1 hour, local xenon redistribution during this short interval will no^ be significant, and operation may proceed without further restriction. l An alternative to realigning a single misaligned RCCA to the group everage dammd position is to align the remainder of the group to the position of the misa11gned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.65, " Shutdown Bank
. l l
I (continued) WCGS-Mark-up ofNUREG 1431-Bases 3.1 B 3.1 26 5/15M7 L - - ------ --- - -- 1
BASES ACTIONS EL1 (continued Insertion Limits " and LC0 3.1.76, " Control Bank Insertion Limits." The Completion Time of 1 hour gives the operator j sufficient time to adjust the rod positions in an orderly manner. l B.2.1.1 and B.2.1.2 With a misaligr.2d rod. SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit. In many cases, realigning the remainder of the group to the misa11gned rod may not be desirable. For exampic, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully 4n insgrfed and control bank C must be ;r.cVed ir, inserted to approximately 100 to-H5 steps id otderJo; maintain. proper; overlap. Power ope _ ration may continue with one RCCA DEEMBM@, -l@ '-'T (,typpibrep but misaligned, provided that SDM is verifEwithin 1 hour. The Completion Timo of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration. B.2.2. B.2.3. B.2.4 _3.2.5. and B.2.6 For continued operation with a misa11gned rod, RTP r,eactorlpoWer must be reduced, SDM must periodiolly be verified within limits, hot channel factors (Fo(Z) and FL) Mt be verified within limits, and the safety analyses must be re evaluated to confirm continued operation is permissible. Reduction of power to 75% RTP ensures that local LHR increases due to a misa11gned RCCA will not cause the core design criteria to be exceeded (Ref. 47+. The Completion Time of 2 hours gives I l (continued) WCGS Mark-up ofNUREG-1431-Bases 3.1 B 3.1 27 S/15/97
ADDITIONAL INFORMATION COVER SHEET ( ADDITIONAL INFORMATION NO: O 3.1-16 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.1.4 Red '3roup Alignment Limits (All FLOG Plants) ITS 3.1.4 Bases Generic Changes Comment: Generic Bases changes need to be discussed / justified. For example, the Bases Background discussion on the DRPI system has been revised and needs to be explained. FLOG RESPONSE: As discussed during a telecon with NRC Staff on June 25,1998, the scope of this RAI will be limited to the ITS 3.14 Background Bases. Changes fallinto one of five categories:
- 1. Specification re-numbering;
- 2. Inclusion of shutdown rods;
- 3. Addition of plant-specific design information (e.g., number of control banks and shutdown banks);
- 4. Editorial corrections (e.g., the correct title for GDC-26); and
- 5. Changes to the last paragraph.
Changes to the last paragraph were made since it was felt that this text went beyond the level of detail required for the ITS Bases. Coil spacing dimensions are not critical to operator understanding of this system. In addition, statements in the last paragraph in the ISTS concerning position indication accuracies are incorrect. ATTACHED PAGES: None l
i l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-19 APPLICABILITY: WC, CA l REQUEST: ITS 3.1.7 Rod Position Indication CTS 3.1.3.2 Position Indication Systems - Operating (Wolf Creek & Callaway) DOC 13-05-A & 13-09-LS-23 & 13-06-A JFD 3.1-7 & 3.1-12 Comment: The ITS retains Conditions and associated Required Actions from the CTS addressing more than one inoperable digital rod position indicator (DRPI) per group, which is not addressed in the STS. However, not all associated CTS Required Actions have been retained in the ITS; the Required Actions to take manual control of the rods and to record reactor coolant temperature every hour have not been retained. These actions, in one case affect rod movement and in the other case provide an indication that the rod (s) position may have changed, and therefore have a bearing on SDM and therefore should not be deleted if the overall condition of more than DRPI per group is inoperable is retained. Either retain the CTS requirements completely, adopt the STS requirements, or provide a better justification for the ITS proposals. The STS wording of l the note permitting separate condition entry should be retained with the STS Conditions and Required Actions. FLOG RESPONSE: The wording of :TS Condition B, with its Required Actions B.1 and B.2, and the change to the Actions Note on separate Condition entry were made pursuant to traveler TSTF-234. TSTF-234 was created based on the Callaway and Wolf Creek CTS; however, Westinghouse and the Westinghouse Owners Group felt that Action Statements b.1.b) and b.1.c) J were unnecessary compensatory actions. The justifications for deleting ' CTS 3.1.3.2 #ction Statements b.1.b) and b.1.c) are discussed in Enclosure 4 under LS-23. In order to capture those justifications under Enclosure 3A, DOC 13-09-LS-23 is revised to add the following:
"The proposed change would delete the Actions to place control rods in manual and record RCS T,y hourly if multiple DRPIs per group are inoperable. Multiple inoperable DRPls, of themselves, have no impact on SDM in MODES 1 and 2 if the control rod positions are verified by alternate means (e.g., movable incore detectors). The requirement to place control rods in manual may not be appropriate in all situations and may be detrimental for load rejection transients unless operator action is assumed to simulate the rod control system in automatic. Accidents l
analyzed using the [ Revised Thermal Design Procetre (RTDP)] assume that the control rods are in (their most limiting mode). Automatic rod movement can accommodate a 10% load rejection. Placing rods in manual may impact the load rejection capability assumed when the P-9 i setpoint was established at 50% RTP. The steam dump system can accommodate a 40% RTP load rejection and with the rod control system in automatic, a 50% RTP load rejection can be accommodated without a reactor trip. While manual operator action can be just as timely as automatic rod control, there is no need to have this limitation in the Technical Specifications. Corrective actions for excessive rod motion are covered under ITS 3.1.7 Condition C. The requirement to monitor and l
l record Ty hourly is unn:cesstry givsn ths availabla indicators and (. alarms, e.g., Tm- T,,, deviation alarm, to alert operators to changing moderator conditions." i ATTACHED PAGES: Encl. 3A 13 4 9 l ,. l
CHANGE NUMBER N2iG DESCRIPTION 13 08 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 1 13 09 LS-23 Current TS Actions b.1.b and b.1.c of LC0 3.1.3.2 are ! deleted. SDM is ensured in MODES 1 and 2 by rod position. ) Multiple inoperable DRPIs will have no impact on SDM in
- MODES I and 2 if the control rod position are verified by alternate means and rod motion is limited consistent with the accident analysis. Deletion of these requirements is consistent with traveler @DCr23rBeG j m. 3.s-act.l
--L e A8-M l WF -250 14-01 -
Not applicable to WCGS. See conversion Comparison Table (Enclosure 3B). 15 01 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 15 02 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 16 01 LS 14 This TS would be revised to apply to shutdown " banks" instead of shutdown " rods"; this is consistent with NUREG-1431. Rev. 1. The current Action Statement permits one rod to be inserted beyond the limits; the proposed ITS l CONDITION A would allow one or more banks to be inserted i beyond the limit. ! 16 02 M The proposed changes to the Action Statement would require l that the shutdown banks be aligned within limits and that SDM be verified or restored. The new Action Statement would extend the time to achieve alignment from 1 to 2 hours as justified in the Bases for ITS 3.1.5. The new Action Statement would establish a Completion Time of 1 hour for verifying and restoring SDN. In the proposed Action Statement, both the realignment and the SDM verification would be required. The current Action Statement provides a 1 hour limit to achieve realignment and effectively applies a 2 hour Completion Time to SDM verification and restoration (which would be performed under the TS for rod group alignment limits). In the current Action Statement, either the realignment or the l SDM verification are required. The current Action Statement could, in some circumstance, allow continued POWER OPERATION with a shutdown rod out of alignment because it was written to apply to individual rods and refers to the rod group alignment specification. The new act%n statement, which applies to shutdown banks, would s..,c permit operation with a shutdown bank outside its WCGS-Description of Changes to CTS 3.1 13 5/15/97 L_
INSERT 3A-13a 0 3.1-19 The proposed change would delete the Actions to place control rods in manual l and record RCS T , hourly if multiple JRPIs per group are inoperable. Multiple inoperable DRPIs, of themselves, have no impact on SDM in MODES 1 and 2 if the control rod positions are verified by alternate means (e.g., movable incore detectors). The requirement to place control rods in manual may not be appropriate in all situations and may be detrimental for load rejection l transients unless operator action is assumed to simulate the rod control l system in automatic. Accidents analyzed using the [ Revised Thermal Design Procedure (RTDP)] assume that the control rods are in [their most limiting mode]. Automatic rod movement can accommodate a 10% load rejection. Placing rods in manual may impact the load rejection capability assumed when the P-9 setpoint was established at 50% RTP. The steam dump system can accommodate a 40% RTP load rejection and with the rod control system in automatic, a 50% RTP load rejection can be accommodated without a reactor trip. While manual l operator action can be just as timely as automatic rod control, there is no ! need to have this limitation in the Technical Specifications. Corrective actions for exce.sive rod motion are covered under ITS 3.1.7 Condition C. The ) requirement to monitor and record T , hourly is unnecessary given the l available indicators and alarms, e.g. , T , - T,,, deviation alarm, to alert operators to changing moderator conditions. l t____._______________.__ _ . _ _ _ _ _ . . _ _ . . . _ . . .
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-23 APPLICABILITY: WC REQUEST: ITS 3.1.6 Control Bank Insertion Limits CTS 3.1.3.6 Control Rod insertion Limits (Wolf Creek) DOC 17-04-LS-8 JFD 3.1-3 Comment: Required Action C.1 in the STS, and in the related actions of the CTS and the TS of the other FLOG Plants, all require the plant to enter Mode 3. The ITS Required Action C.1 requires the plant to be placed in Mode 2 with Keff<1. Maintain consistency with the STS and the CTS. FLOG RESPONSE: Wolf Creek withdraws the proposed change to revise ITS Required Action C.1. The Bases for ITS LCO 3.0.2 indicate that if a Required Action is not comple??d within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. In accordance with plant procedures, the plant is brought to MODE 3 to allow for more stable plant conditions prior to resumption of power operation. The proposed change was being made to maintain consistency between the Mode of Applicability and the Required Action. The proposed change was based on TSTF-26 which has been approved by the NRC. TSTF-26 is applicable to STS 3.4.2, RCS Minimum Temperature for Criticality, and revised Required Action A.1 to state: "Be in MODE 2 with Kon< 1.0." The Mode of Applicability for STS 3.4.2 is MODE 1, MODE 2 with K,n 21.0. The basis for the change was to l l maintain consistency between the Mode of Applicability and the Required Action which requires the Mode of Applicability to be exited. In November 1996, a proposed traveler was submitted to the WOG MERITS Mini-Group which proposed changing STS 3.1.7 Required Action l C.1 to state "Be in MODE 2 with Kon< 1.0." The proposed traveler was I rejected by the Mini-G.*oup. Additionally, another licensee oroposed this same change in their conversion application and the change was not j
~
questioned by the staff. ATTACHED PAGES: Encl. 2 1-14 Encl. 3A 15 Encl. 3B 10 Encl. 4 - 1. new LS-21 Encl. 5A 3.1-15 Enci5B B 3.1-41 Encl.6A 1 Encl. 6B 1
l REACTMTY CONTROL SYSTEMS l l CONTROL ROD INSERTION LIM.IIS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be F--!!:f M ph ; 9-$Oin the) insertion,[ sequence) (and overlap limits)es specified in the CORE OPERATING LIMITS REPORT (COLR). $@""741MR **N APPLICABILITY: MODES nd kxcept for surveillance testing pursuant to) EjM034fM?j*l
*hmMkindd
[8 specification 4.1.3.1.2) ACTION With the control banks inserted beyond the insertion limita(or not within sequence) gggygygg fand overlap ilm"'--- *" : *- m' " - - - - ^8---~--"---- - ^ - - - " - - - - ' * - wwcasmisem r.,u__*-- 4 g
.sw.a 4..Wa9 l a. Within 1 hour, venfy tha64he SHUTDOWN MARGIN isipeatesthew l
--" ' *- 1.?% 9_" 3o be w6 thin limits provided in the COLR,)
$$M4}M'QJ NNdWW or initiate borabon until the SHUTDOWN MARGIN is restored leper'ec. Or ?-"r' te 1.3% 9_^ ,twithin limit), nd
- b. Restore the control banks to within the limits within 2 hours, or
- c. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specrfied in the COLR, or r- ,rn
- d. Be in -:::" HOT ST^N ' [g pEI)4jpf % ji/J) withinghe next)6 hours. %3R r f @ I~'
{l"7-03-Lb-()
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kt led @T STAklDB SURVEILLANCE REQUIREMENTS udiste. --{ ca 3.t-oo Q 4.1. 6.1 The position of each control bank shall be determined to be within
'h :r;. ., limits at least once per 12 hour: r-^ ;t it; S .: ! ':. :!: ; 512il40Gyd
-T.; . P.d f r;- -- U " ": :^:r !: 1 0---d'r, 'ern "0-*/ 'M !^ 'S/:t2! ^****C' i d ; :?? :x ' '---t :x: p- ' M"-
4.1.3.6.2 When in Mode 2 with K.e less than 1, verify that the predicted entical control rod position is within insertion limits within 4 hours prior to achieving reactor criticality. I(New) Verify sequence and overlap limits specified in the COLR are met for e ' 17-01-M '; gcontrol banks not fully withdrawn from the core at least once per 12 hours. ""*"^^ " "'J
*S:: !;r!:' Tet 9::;f:x
- er*::t: : 3.10.2 ed 3.10.2. STD346A5"
#With K,g greater than or equal to 1. NndnNW i
l l l WOLF CREEK - UNIT 1 3/4 1-14 Amendment No. 64,89 Markup ofCTS 3M.] S/15n7 l w-_---___-__._--_._-- - _
CHANGE Hl.tEER RSE DESCRIPTION 17 02 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). _ rio.1u-231 17 03 - L6 21 Not pli _1 to WBGSMee Coffvepsfon_ f4mparisor)Eabij clo re ). t545Estr 3A-15a f - Wolf Creek CT ction 3 1.3.6.d equir s that ith the control b s not wi in limi .t unit in STANDB MDDE 3) . thin 6 ur Cond on C IT .1.6 r res that unit p ed i E2 th ,< ithin 6. urs if uir ction oci comple n time no met. CT as n fied to re t this fa ed co 1 tion i req es tha he t be p ed a in ch L s no pply. This c s acc able nc aci the in 2 wit , 1.0 ' con e ith Appli ility ensu t r ist utio hat d res in fuel c1 di ailu i ev of a . lo of fl . I eje r or r ac nts uirin ermin on by i an RTS p ction e prev ed. sist with NUR -4 . Rev. this nge a add addit al 2 l hours the owed o age t to e the of icabili for t LCO. i I WCGS-Description of Changes to CTS 3.1 IS S/15/97 i
INSERT 3A-15 0 3.1-23 Consistent with NUREG-1431 Rev. 1. this change adds an additional two hours to the allowed outage time to be in MODE 3. l {
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NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC) CONTENTS I. Organization ........................................................... 2 II. Descri ption of NSHC Eval uati ons. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l l III. Generic No Significant Hazards Considerations t I l
- A . Administrative Changes.............................................. 5 l; l
I L R Relecated Technical Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l LG Less Restrictive (Moving Information Out of the Techni cal Speci fications) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 H - More Restrictive Requirements....................................... 12 i t i IV. Specific No Significant Hazards Considerations LS l LS 1.................................................................... 15 LS-2.................................................................... 17 LS 3.................................................................... 19 LS 4.................................................................... 21 LS 5.................................................................... 23 LS 6.................................................................... 25 LS 7.......................................................... = _ . 28 LS.8......................................................... .M...u.s.t.a.... LS , 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 y o s. i - n l LS 10................................................................... 38 LS 11..............................................................Not Used LS 12................................................................... 40 LS 13................................................................... 43 LS 14................................................................... 45 LS 15.................................................................... 47 LS 16. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not a ppl i cabl e LS 17........................................................Not applicable LS 18................................................................... 49 LS .19. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not a ppl i cabl e LS 20........................................... ._ - ...Not applicable l LS 21......................................... " M......_.A.... . ..yy ". " D.^ n. .n-231 LS 22................................................................... 51 LS 23................._._.... ~ ~ 1 ..... . . . V. Q.5-24 Generic Technical NSHCs _~ msa.cr 4.-b TR s.t -oca l . . . . . . . . . . . . . . l ! TR 2.................................................................... 55 TR 3.................................................................... 57 WCGS-NSHCs-CTS 3M.1 1 5/1S/97
INSERT 4-a 0 3.1-23 i IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-21 10 CFR 50.92 EVALUATION FOR
. TECHNICAL CHANGES THAT' IMPOSE LESS RESTRICTIVE RE0VIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS l The proposed. change would revise the shutdown requirement for control rod banks not within limits from "Be in at least HOT STANDBY within 6 hours" to "Be in at least HOT STANDBY within the next 6 hours". This change would' provide additional time to place the plant in MODE 3 because the 6 hour limit L would not ' commence until after the time periods allowed for completing other actions, such as restoring the control bank position within 2 hours. The current TS require the 6-hour period to commence at the time the LCO is not met. The proposed change is in accordance with NUREG-1431, Revision 1.
i This proposed TS change has been evaluated and it has been detcrmi9ed that it involves no significant hazards consideration. This determination has uten
-performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:
1 "The Commission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating ifcense for a facility licensed under 50.21 (b) or 50.22 or for a testing l facility involves no significant hazards consideration, if operation of l the facility in accordance with the proposed amendment would not: 1 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or l 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. "
l The following evaluation is provided for the three categories of the ! significant hazards consideration standards: l l
- 1. Does the change involve a significant increase in the probability
.or consequences of an accident previously evaluated?
The control rod insertion LC0 is intended to assure that the control rods are in position to mitigate accidents or transients that depend on a reactor trip and that reactor power. distribution is in accordance with the assumptions used in accident analyses. If control rods are outside of specified limits, the plant should be placed in a MODE for which the LCO is not applicable. i.e.. , MODE 3. As noted'in the Bases of NUREG-1431, Rev. 1, six hours is a 1 1 a
IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATION NSHC LS-21 (continued) reasonable time, based on operating experience, for reaching the required MODE from full power conditions, in an orderly manner without challenging plant systems. l The proposed shutdown requirement completion time change would result in an extension of time to achieve MODE 3 from 6 hours (per current TS) to a maximum i of 8 hours (2 hours to attempt to complete other mitigating actions followed by 6 hours to achieve MODE 3). The proposed change would not change the plant design or operations such that an accident or transient could be initiated. l By allowing a shutdown time based on operating experience, the change would
- reduce the chances of an operator error or challenge to plant systems that l could result from the more restrictive requirements in the current TS. Thus, l the change would have no adverse effect on the probability of occurrence of an accident.
The proposed change would not affect the method of operation of plant systems and involves only the time requirement to achieve a reactor shutdown when the l control rods are not within limits. The probability that an accident would occur during the 2-hour time extension allowed by the proposed change would be l extremely small. Also, during the 2-hour period, the other required actions are being performed. These include returning control rods to within limits, restoration or SHUTDOWN MARGIN as necessary, and reduction in reactor power. Therefore, as time in the action statement passes, the plant is being brought closer to conditions for which the accident assumptions are valid. Thus, the l proposed change would have a negligible effect on the consequences of accidents previously analyzed. 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 l of accident from any accident previously evaluated? Operation in accordance with the proposed change would not introduce any new failure modes for plant systems and componenets. Only the duration of l operation in the action statement is affected. The procedures for operating l plant equipment and the configuration of plant equipment are not affected. 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?
l The margins of safety involved with this proposed change are those associated with accidents that rely on control rod position to assure reactor trip effectiveness and to assure power distribution limits are maintained. When the rod position is not within limits, the LC0 initiates the corrective action to restore the margins by specifying mitigating actions. While the margins of I
! IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATION NSHC LS-21 (continued) safety may be affected by failure to meet the LCO, the additional two hours to l achieve MODE 3 allowed by the proposed change has no effect on them. The ! . control rod insertion limits are not changed nor are there any changes to l accident analysis assumptions, methodologies, credited protection / mitigation equipment, or event acceptance criteria. 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
' the overpower limits, DNBR limits. Fa, F H. LOCA PCT, peak local power density, or any other margin of safety.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION l Based on the above evaluation, it is concluded that the activities associated l with NSHC "LS-21" resulting from the conversion to the improved TS format j satisfy the no significant hazards consideration standards of 10 CFR 50.92(c), and accordingly, a no significant hazards consideration finding is justified. l ! i l I ( i l _ _ _ _ _ _ _ _ _ _ _ . _ . - - _ . _ _ _ _ _ _ _ _ _ _ - - _ - - - - - - - _ . . - - - - - _ - - _ - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - ~ - --
i Control Bank Insertion Lisits 3.1.76 gIE9Fi
- i i
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Control bank sequence or B.1.1 Verify SC" O a [1.G 1 hour overlap limits not met. f 1.M ok/k.tYbeTrfthin $$$!$18j ; thelteits:~provided1D tLie;UXR; Gl . B.1.2 Initiate boration to 1 hour restore SDM to within limit. AND 4 B.2 Restore control bank 2 hours sequence and overlap to within limits. I C. Required Action and C.1 y Be in MODE ~ p s.i 23 6 hours -w l associated Completion p. (d$)#i3h { Time not met. - v' ! 1 l 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.76.1 Verify estimated critical control bank position Within 4 hours is within the limits specified in the COLR. prior to achieving criticality (continued) I WCGS-Mark-up ofNUREG-1431-ITS 3.1 3.1 15 S/lS/97
Control Bank Instrtion Lisits ! B 3.1.76 j BASES l ACTIONS A.1.1. A.1.2. A.2. B.1.1. B.1.2. and B.2 (continued) l Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits. . Operation beyond the LCO limits is allowed for a short time l period in order to take conservative action because the i simultaneous occurrence of either a LOCA, loss' o'f flow accident, l ejected rod accident, or other accident during this short time ! period, together with an inadequate power distribution or l reactivity capability, has an acceptably low probability. The allowed Completion Time of 2 hours for restoring the banks to within the insertion, sequence, and overlap limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time. F haier M appbrFNioesinotTregsdire"gtgrMBIeLetqcas3sjl Lag ars strchemwTaruverl.ap:wmits7arerentague_ C.J If Required Actions A.1 and A.2, or B.1 and B.2 cannot be N r=nleted within the associated _ Completion Times, the plant must b be brought toMIBE-3(IUlGedEWL,M. where the LCO__is not
$ 3123 \
sg51UNetJ!iWlB!1oKiitintN6agmiti!Mk!!E l The allowed Completion Time of 6 hours is reasonable, based on l operating experience, for reaching the required H0DE from full l power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.76.1 i REQUIREMENTS , This Surveillance is required to ensure that the reactor does not ! achieve criticality with the control banks below their insertion limits. 1 (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1-41 5/158 7
DIFFERENCES FROM NUREG-1431 Section 3.1 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431. Revision 1, to make them plant specific or to incorporate generic changes resulting from the Industry /NRC generic change process. The change numbers are referenced directly from the NUREG-1431 mark ups. For Enclosures 3A, 38, 4, 6A, and 68, text in brackets "[ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcasuiittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUfBER JUSTIFICATION g 3,,4 3.1 1 In accordance with TSTF 9, this change would relocate the specified limit for Shut in (SDM) from the ISTS to the COLR. This change occurs in several specifications including the Specification for SDM and those specifications with ACTIONS that require verifying SDM within limits. 3.1 2 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). Q3.1-23 3.1-3 Wolf Cr ITS L 3.1. Requir Acti C.1 1 rev rom ~ in . Be i E th < 1.0 " icabi y of 3. . is 1a 2 wit n m .. a tion. he ses scus n indi ed t equir Act s' ntent s to pl the it in the is ta icab : this 1d 2 ith 1.0. his nge sim ar to the (cha propos in tr eler 26. M (Astd.. 3.1-4 SR 3.1.4.2 of NUREG 1431. Rev. I would be deleted. In accordance with TSTF 13,yhrt/Z)the intent of this SR is only to determineys., oor,] the next frequency for SR 3.1.4.3. Perfonnance of SR 3.1.4.2 is ! not necessary to assure that the LCO is met. SR 3.1.4.3 fulfills ; that purpose. Therefore. SR 3.1.4.2 may be deleted. In addition, the Note in the Frequency column of SR 3.1.4.2 would be moved to become Note 1 in the Surveillance coltan of SR 3.1.4.3. This is for clarification purposes. As discu:: sed in CN 3.19, section re numbering results in SR 3.1.4.3 of NUREG 1431, Rev.1 becoming ITS SR 3.1.3.2. 3.1 5 Per current TS [3.1.3.1], the words "with all" have been removed from ITS LCO 3.1.4. This is a clarification that ensures the proper interpretation of the LCO. The change makes it clear that only one channel of DRPI is necessary to meet the alignment accuracy requirement of the LCO. With the word "all" in the WCGS-Differencesfrom NUREG-1431-ITS 3.1 1 5/15/97
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i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-24 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.1.4 Rod Group Alignment Limits (All FLOG Plants) JFD 3.1-5 & 3.1-6 Comment: Rewording of LCO and Condition A approved, contingent upon OG resubmittal of change request TSTF-107 (revision) as discussed with TSTF. FLOG RESPONSE: See the response to Comment Number 3.1-15. The FLOG has incorporated TSTF-107, Revision 1. ATTACHED PAGES: None l l
ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: O 3.1-25 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 3.1.4 Rod Group Alignment Limits (All FLOG Plants) JFD 3.1-16 l i Comment: Inclusion of SR 3.2.1.2 to Required Action B.2.4 is approved; ensure OG submit WOG-105 as a TSTF change request. ' FLOG RESPONSE: At the June 23-24,1998 meeting of the Westinghouse Owners Group MERITS Mini-Group, traveler WOG-105 was discussed. The remaining I action on this trave!sr was assigned to Westinghouse to expand this change to also apply to ISTS 3.2.1 A, "Fo (Z) (F,, Methodology)." However, this additional work has no impact on the manner in which the FLOG has ! incorporated this traveler's additional restriction. The TSTF will be I submitted to NRC expeditiously. ATTACHED PAGES: Encl. 5A Traveler Status page i l l l
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS SpecialTest Exceptions 3.1.10 is retained and renumbered as 3.1.8, consistent with this traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. I in s.s.cor l TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. (TSTF-10%D Incorporated 3.1-6 . le J.:-is' 1 TSTF-108, N: :;;;p . :;d -NA- Not NRC approved as.es R** 1 __ _ lacorPor*ted _ _ ll-_2d __ _ . __ N. Ta s.i-esil l
+,
TSTF-110 Incorporated 3.1-10 ! Rev. f- N "PPr* - W i-#4 ! TSTF-136 Incorporated 3.1-9,3.1-15 (NRc pM-lTe.it- cor.j l TSTF-141 Not incorporated NA Disagree with change; ; traveler issued after cutoff i date TSTF-142 N5 ;G.Af;3 ^ -NA- T.Tc;'d ; ;d ;;;r# ** p 4Pt agrah Incorporstel-. _ 3. i - 2.2. _ ! ( Incorporated 3.1-7 (WOG3 D [ Incorporated _ E 3.1-16 'Z.e.,, GD/@2 8- as I
~
]7R.3.l-004,]
i 5/15/97
1 l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.1-27 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 3.1.1 Shutdown Margin (All FLOG Plants) l l JFD 3.1-18 l Comment: This modification adds a Mode change restriction from Mode 6 to Mode 5, as discussed in CN 1-02-LS-1 of 3.0. The discussion provided is inadequate to evaluate the necessity of the mode change restriction. In general, throughout the submittal, I justifications for notes prohibiting mode changes are inadequate. Provide j explanations / justifications that present specific conditions that would necessitate the note. FLOG RESPONSE: A Reviewer's Note in STS LCO 3.0.4 states: "LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis, the 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 3.1-18 has been revised to incorporate additional justification from NSHC LS-1 from Enclosure 4 of the Section 3.0 package (Attachment No. 6). JFD 3.1-18 has been revised to 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 3.1.1 was modified by a Note stating: "While this LCO is not met, entry into MODE 5 from MODE 6 is not permitted." Entering MODE 5 without SDM limits met implies that boron concentration in MODE 6 is not met. Under these conditions, a transition to MODE 5 should not be attempted until MODE 5 SDM limits are met. Inadvertent boron dilution events are precluded in MODE 6 via administrative controls (that close the dilution source valves), whereas dilution events are not [ physically precluded) in MODE 5. Therefore, the transition from MODE 6 to MODE 5 should not be allowed in the SDM initial condition for a MODE 5 dilution event is not met." l ATTACHED PAGES: Encl.6A 4
CHANGE NUleER JUSTIFICATION l l
' 3.1 17 Consistent with current TS LC0 3.1.3.2 and the wording of ITS l 3.1.7 Conditions A and B. ITS 3.1.7 Condition C is clarified to state that the inoperable position indicators are inoperable DRPI's. I 3.1 18 A MODE change restriction has been added to ITS 3.1.1 in the LCO Applicability, per the matrix discussed in CN 102 LS 1 of the 3.0 package. (See the LS-1 NSHC in the CTS Section 3/4.0, ITS Section 3.0 package).
{sE'#.T j q s.t -M 3.1 19 Not used. 3.1-20 Consistent with current TS 3/4.10.3 " Physics Tests," ITS LCO 3.1.8 and its Condition C and SR 3.1.8.2 are modified to refer to
" operating" RCS loops. Adopting the current TS is acceptable since valid T,,, measurements are not obtainable for a non-operating loop.
.3.l-2.1 iWSEst1 GA-% j TM a l.co 1
- 2.~2. "TIEL 3. I -003 1
i l i l l WCGS-Differencesfrom NUREG-1431-ITS 3.1 4 S/158 7
INSERT 6A-4a 0 3.1-27 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 3.1.1 was modified by a Note stating: "While this LC0 is not met, entry into MODE 5 from MODE 6 is not permitted." Entering MODE 5 without SDM limits met implies that boron concentration in MODE 6 is not met. Under these conditions, a transition to MODE 5 should not be attempted until MODE 5 SDM limits are met. Inadvertent boron dilution events are precluded in MODE 6 via administrative controls [that close the dilution source valves], whereas dilution events are not ( [ physically precluded] in MODE 5. Therefore, the transition from MODE 6 to I MODE 5 should not be allowed in the SDM initial condition for a MODE 5 ) dilution event is not met. I INSERT 6A-4b TR 3.1-001 1 3.1-21 The ISTS SR 3.1.8.1 requirement to perform a CHANNEL OPERATIONAL J TEST (C0T) on the intermediate and power range NIS channels within 12 hours prior to initiating PHYSICS TESTS is revised to delete phrase "within 12 hours." COT testing is performed on these channels prior to reactor startup per LCO 3.3.1. This change is consistent with traveler TSTF-108. INSERT 6A-4c TR 3.1-003 3.1-22 The Completion Times for ITS 3.1.2, Required Actions A.1 and A.2 are increased from 72 hours to 7 days, consistent with traveler TSTF-142. 1
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.1-28 APPLICABILITY: DC, CP, WC, CA REQUEST: Relocated Specifications (All FLOG Plants) Comment: Comanche Peak, Wolf Creek, and Callaway have not provided relocated screening evaluations / forms for any of their specifications relocated to licensee controlled documents. Diablo Canyon has not provided relocated screening forms for all of their specifications relocated to licensee controlled documents. Provide necessary relocation screening evaluations / forms. FLOG RESPONSE: All relocated specifications have been provided the necessary ielocation screening evaluations / forms which are contained in Attachment 21. For Callaway and Wolf Creek, Section 3.1 specifications were previously relocated by Amendment No.103 and 89 respectively. Therefore, none of the relocation DOCS apply to Callaway and Wolf Creek and this question is not applicable to these plants. ATTACHED PAGES: None i
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.1-001 APPLICABILITY: WC, CA REQUEST: Retain words "the insertion" in CTS SR 4.1.3.6.1. Revise ITS Bases for SR 3.1.2.1 and 3.1.8.4 to reflect changes made to ITS Bases for SR 3.1.1.1 (re: shutdown rod position and, for Callaway, boron-10 depletion). ATTACHED PAGES: Encl. 2 1-14 Encl. 58 B 3.1-13 1 I
REACTMTV CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be !' F:f in ph; fr[f.^2.:n the) insertion.[ sequence) {~g*
~"* 1s01MJ Cnd overtap limits)as specified in the CORE OPERATING LIMITS REPORT (COLR).
APPLICABILITY: MODES xcept for surveillance testing pursuant to) [5pecafication 4.1.3.1.2) nd {W.0A] ] ACTION With the control banks inserted beyond the insertion limita(or not within sequence) p - ' pr [and overlap lim" -- - * ' ' ' "^' " - ' '-- -"- " ^ ^
- - - - ' * - q f . ^ *WM S;::* ^ ' ?.2.1.2:
- a. VAthin i hour, verify theHhe SHUTOOWN MARGIN !: z _ t:n :: g p*" " @P M :84@*.l('""
g
--" ' S 1.?% fl be within limits provided in the COLR,) S or initiate boration until the SHUTDOWN MARGIN is restored 6 :: ':-". - -" - 1.2% ff,lwithin limit). nd
- b. Restore the control banks to within the limits within 2 hours, or
- c. Reduce THERMAL POWER wdhin 2 hours to less than or equal to that ,
frachon of RATED THERMAL POWER which is allowed by the bank ' position using the inserten limits specified in the COLR or-- r- i rn
- d. Bein d'- '"^T ST.^."99 g)4jpfj%ji/J) 3, wNhinghe nentl6 hours. L W M- - - . @3 I~Ib
{l'1-03-LS-J}. QlodHarSTAuDB SURVEILLANCE REQUIREMENTS uvwle.le.te. --{ cA 3.1-oo l]_ 4.13.6_.1 The position of each control bank shall be determined to be within _ fn::Climits at least once per 12 hour: _ _;: ddn;'l- .: in::. ;;'-
- 2: . -- PM fr:;rl : U-f:n?.: !:! :;:- t' ,9r ;;f;"-!dMte' d ;- ^' :: d ! :-' : :: p-- ' 50" .
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4.1.3.6.2 When in Mode 2 with K.,less than 1, verify that the predicted 4 l critical control rod poorbon is within insertion limits within 4 hours prior ! ! 18 achieving reactor cnbcality. i J l l-ONew) Verify sequence and overlap limits specified in the COLR are met forl control banks not fully withdrawn from the core at least once per 12 hours. } {g'/p
* }}
I C?-^ ?;-i ' T^ ' E. :- ;'l- .: I; -P : ^ --- 3.10.2 ed 3.10.2.
#With K,g greater than or equal to 1. M l
1 l i WOLF CREEK- UNIT 1 3/4 1-14 Amendment No. 64,89 Markup ofCTS3M.] M587
Core Reactivity B 3.1.32 BASES l SURVE LLANCE SR 3.1.32.1 (continued) - l REQUIREMENTS **d 8h"N* ""~# 'l l including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering H00E I at an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (iinecessaly) of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering H0DE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly. REFERENCES 1. 10 CFR 50, Appendix A. GDC 26. GDC 28, and GDC 29.
- 2. FSAR USAR, Chapter E153 i
WCG5'-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 13 5/158 7
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 3.1-001 APPLICABILITY: DC, CP, WC, CA REQUEST: Incorporate NRC-approved traveler TSTF-108 Revision 1 to delete the words "within 12 hours" from the Frequency of CTS SR 4.10.3.2 and ITS SR 3.1.8.1. 1 l ATTACHED PAGES: l l Attachment No.16 - CTS 3/4.10 Encl. 2 10-2 Encl. 3A 2 Encl. 3B 1 Encl. 4 1, (new LS-1) Attachment No. 7 - CTS 3/4.1, ITS 3.1 Encl. 5A Traveler Status page,3.1-21 Encl. 5B B 3.1-57 Encl. 6A 4 Encl. 6B 3 l l l
SPECIAL TEST EXCEPTIONS 3/410 3 PHYSICS TESTS g LIMITING CONDITION FOR OPERATION 3.10.3 The limdations of Specifications 3.1.1.3,3.1.1.4. 3.1.3.1,3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
- k. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER.
'_%,_ _? :'...'.'. . _ _ _the.mC"_""_ _^"._,E ' ,f.1_
- b. 5 "_- - ^--_m___ .
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. . , m,ven :": i,
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phNAhDil is within the timkhspecified In the COha'd n . t-M))
- c. The Reactor CMat System lowest operating loop temperature (Ty is greater than or equal to 541*F.
APPLICABILITY : MODE 2(uring PHYSICS TESTS] {344]- ACTION:
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers. j l b. With a Reactor Coolant System operating loop temperature (Ty less inen 541*F, restore T to within its limit within 15 minutes or be in at least HOT STANEY within the next 15 minutes.
now) With SDM not within limit, within 15 minutes, initiate boration to restore Y3 SDM to within limit and, within 1 hour, suspend PHYSICS TESTS
- 3 03-M M8 7 exceptione.
SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than er equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an i ANALOG CHANNEL OPERATIONAL TEST @pfzy J -otr - LS-prior to initiating PHYSICS TESTS.
= -QTA 3.1 -80 / l 4.10.3.3 The Reactor Coolant System temperature (Ty shall be determined to be greater than or equal to 541*F at least once per 30 minutes during PHYSICS TESTS.
(NEW) Verify SDM to be within the limits specified in the COLR at least once per 24 3-01 M-.] ! WOLF CREEK - UNIT 1 3/4 10-2 Amendment No. 89 Mark-up ofCTS3M.IO S/lS/97 i
CHANGE NUMBER HSBC DESCRIPTION
+
3 03 Not used. 3 04 A The applicability statement would be changed to be more consistent with operation for testing purposes. The proposed change is consistent with NUREG 1431, Rev. 1, and m -does _not res_ ult in any changes t_o technical requirements. l (3-05 Lb-I _ _ ms Gar a A .zA}.f ne. .v.t o o I / 4 01 M Special Test Exception [LC0 3.10.4] would be deleted. This specification allows the suspending of requirements of one or more LCOs (depending on plant spec 1Nc current TS) under certain conditions. Elimination of the special test exceptions is justified either because their elimination would be consistent with NUREG 1431, Rev.1, l or because the applicable tests are performed only during initial plant startup and are no longer needed. Therefore, because the exceptions applicable to each LCO eddressed by [LCO 3.10.4] are no longer required [LCO 3.10.4] may be eliminated. This change is acceptable l because it imposes more stringent requirements (i.e., eliminating exceptions). The elimination of the STEs has no adverse impact on the health and safety of the public. 5 01 R Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 38). 5 02 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). ! I l i WCGS-Description of Changes to CTS 3M.10 2 S/15/97
l l INSERT 3A-2a TR 3.1-001 3-05 LS-1 The current SR requiring the performance of [an ANALOG) i CHANNEL OPERATIONAL TEST on each intermediate and power range NIS channel within 12 hours prior to initiating ) PHYSICS TESTS is revised to delete the phrase "within 12 I hours." Current TS LC0 3.3.1. Reactor Trip System (RTS) Instrumentation, requires the performance of an [A] COT on ; the power range low setpoint and intermediate range NIS channels prior to each reactor startup, if not performed , within the previous 31 days (revised to 92 days in the conversion to ITS 3.3). These RTS SRs must be performed prior to entering the LCO 3.3.1 Applicabilities for these { RTS trip functions since there are no CTS SR 4.0.4 exceptions. Current SR 4.10.3.2 requires an arbitrary estimate of when the plant is within 12 hours of intiating i PHYSICS TESTING. This has no basis from the accident analyses, which are satisfied as long as the surveillance are current prior to entering plant MODES where these trip , functions provide protection. When these surveillance are current, they have previously been determined to remain valid for 92 days. The initiation of PHYSICS TESTING does not impact the ability of the channels to perform their required function, does not affect the trip setpoints or trip capability of these channels and does not invalidate the previous surveillance. This change is consistent with traveler TSTF-108. 1 l
_ a W muT / _ . P A A oS i S L - vs L ei s s s s A orh e e e A e C - N pt Y Y Y N Y d 9 e 8t e tl ne K ed m r dy . nl e sE Y T T muT A oS i I F - vs L L ei s s s s I O orh e e e A e B W N pt Y Y Y N Y _ A _ C _ I L . 0 P 1 P K 4 A A E
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T R R E E B 51 _ S M 0 - N U - S N 3L - I ll
NO SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC)
, CONTENTS
- 1. Orga n i zat i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 II. Description of NSHC Evaluations.......................................... 3 III. Generic No Significant Hazards Considerations "A" Admi ni strati ve Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 "R" Rel ocated Techni cal Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 "LG" Less Restrictive (Moving Information Out of the Techni cal Speci fi cations) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 "H" - More Restrictive Requirements...................................... 12 IV. Specific No Significant Hazards Considerations "LS"
-4iene- LS-1 tit.3.1-ooi}.
1 l l [ \ WCGS-NSHCs - CTS 3M.10 1 S/2507
INSERT 4 TR 3.1-001 IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-1 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The current SR requiring the performance of [an ANALOG] CHANNEL OPERATIONAL TEST on each intermediate and power range NIS channel within 12 hours prior to initiating PHYSICS TESTS is revised to delete the phrase "within 12 hours." This change is consistent with traveler TSTF-108. 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 l l
- 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: i
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
l Overall protection system performance will remain within the bounds of the , i previously performed accident analyses since no hardware changes are proposed. l Current TS LC0 3.3.1 Reactor Trip System (RTS) Instrumentation, requires the performance of an [A] COT on the power range low setpoint and intermediate range NIS channels prior to each reactor startup, if not performed within the
~
previous 31 days (revised to 92 days in the conversion to ITS 3.3). These RTS SRs must be performed prior to entering the LC0 3.3.1 Applicabilities for these RTS trip functions since there are no CTS SR 4.0.4 exceptions. Current i SR 4.10.3.2 requires an arbitrary estimate of when the plant is within 12 hours of intiating PHYSICS TESTING. This has no basis from the accident ' analyses, which are satisfied as long as the surveillance are current prior to entering plant MODES where these trip functions provide protection. When these surveillance are current, they have previously been determined to
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-1 i (continued) l recain valid for 92 days. The initiation of PHYSICS TESTING does not impact the ability of the channels to perform their required function, does not affect the trip setpoints or trip capability of these channels and does not invalidate the previsous surveillance. 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 or different kind of bccident from any accident previously evaluated?
There are no hardware changes nor are there any changes in the method by which any safety-related piant system performs its safety function. The change to the SR will not affect the normal method of plant operation. 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. l 3 Does this change involve a significant reduction in a margin of safety? l I The proposed change does not affect the acceptance criteria, analysis assumptions, methodologies, or credited equipment 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. l NO SIGNIFICANT HAZARDS CONSIDERATION DETERMIN?, TION l i Based on the above evaluation, it is concluded that the activities associated l with NSHC "LS-1" resulting from the conversion to the improved TS format j sati.sfy the no significant hazards consideration standards of 10 CFR 50.92(c); I dnd aCCordingly, a no significant hazards consideration finding is justified. 4 I
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 ' TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS SpecialTest Exceptions 3.1.10 is retained and renumbered as 3.1.8, consistent with this traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. I'M 8 8 865 l TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. 1 (TSTF-107,b.- 1) Incorporated 3.1-6 . le 2.'s -# r 1 ) N:: id:: p-.:[
~
TStF-iO8, . -NA-- Not NRC approved nW Rev.1 - incorporated _ -- - --.51-2.1 _ _ = N. - Ta s.i-o.i1 I TSTF-110 Incorporated 3.1-10 w l Rev. f. N a#" - N I~# ! TSTF-136 Incorporated 3.1-9,3.1-15 (pRc eM-tva s.t- cord TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoff date TSTF-142 N:U ..G.Je2
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~
j 'IR. 3. 8 - Colo] 5/15/97
l l I PHYSICS TESTS Exceptions-MODE 2 3.1.H8 UIN58N
!$$3!1@@
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCS lowest operat_ing loop average C.1 Restore RCS lowest naarating loop average 15 minutes ggg temperature not within temperature to within limit. limit. D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met.
, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.%8.1 Perform a CHANNEL OPERATIONAL TEST on power "ithir. ggg range and intermediate range channels per 12 tsr:;
fior to SREMZindlab]er33p3 Tw,si-cei L J initiation of PHYSICS TESTS SR 3.1.168.2 Verify the RCS lowest gperatitgCloop average 30 minutes e-X:J3 temperature is 25M 541*F. ({"f" gg SR.~i3318,1 yeMTL GMjtL !gLER~,JsjfMR; LM WN$% SR 3.1.%8.34 Verify SDM is .1.'.A ok/k. withJn]1mits 24 hours M Ed pnovidedlinthe_COjR. ' WCGS-hfark-up ofNUREG-1431-ITS 3.1 3.1-21 S/1587
PHYSICS TEST Exceptiohns H0DE 2 B 3.1.198 i l BASES
. l ACTIONS lL1 (continued) the reactor and prevent operation of the reactor outside of its design limits. j C.l When the RCS lowest operatingloops Tm is < S E S41*F the appropriate action is to restore Tm to within its specified limit. )
The allowed Completion Time of 15 minutes provides time for restoring Tm to within limits without allowing the plant to remain in an , i unacceptable condition for an extended period of time. Operation with ! the reactor critical and with an~tterWttig[g1s temperature below ~ 531541*F could violate the assumptions for accidents analyzed in the safety analyses. D.l If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least H0DE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3. L 408.1 REQUIREMENTS The reduired power range and intermediate range neutron detectes must be verified to be OPERABLE in MODE 2 by LC0 3.3.1, " Reactor Trip System (RTS) Instrianentation." A CHANNEL OPERATIONAL TEST is performed on each OPERABLE power range and intermediate range channels Mprior to lnitiation of the PHYSICS TESTS. This will ~
~
ensifre that the RTS is properly aligned to provide the required ree o ore protection du ing the performance _of the PHYSICS TESTS. g g ,u , L ti rt Timi ssuffJcTent before,Anitiati en that PHY CS TEST . insfume tio is
/
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- ~ =-
whe. MSM^ 3 m m c.s,
=
} (continued) FYCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 57 S/2S/97
l Cl%NGE . l NUMBER JUSTIFICATION 3.1-17 Consistent with current TS LC0 3.1.3.2 and the wording of ITS 3.1.7 Conditions A and B, ITS 3.1.7 Condition C is clarified to state that the inoperable position indicators are inoperable , DRPI's. ! 3.1-18 A MODE change restriction has been added to ITS 3.1.1 in the LCO Applicability, per the matrix discussed in CN 102 LS 1 of the 3.0 package. (See the LS 1 NSHC in the CTS Section 3/4.0, ITS l Section 3.0 package). j q 3,g .u-3.1 19 Not used. 4 l 3.1 20 Consistent with current TS 3/4.10.3. Thysics Tests," ITS LC0 I 3.1.8 and its Cendition C and SR 3.1.8.2 are modified to refer to
" operating" RCS loops. Adopting the current TS is acceptable l since valid T,,, measurements are not obtainable for a non-operating loop.
.3.1-2.1 i sEs.T (,A -% -
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.3. l - 22. L user.T (,A - 4c. ,
7R 3. I -003 l WCGS-Differencesfrom NUREG-1431-ITS 3.1 4 S/1.. '
INSERT 6A-4a 0 3.1-27 LCO 3.0.4 has been revised so that changes in MODES or other specificd conditions in the Applicability that are part of a shutdown of the unit shall i 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 previo.usly applicable in all MODES. ITS LCO 3.1.1 was modified by a Note stating: "While this LC0 is not met, entry into MODE 5 from MODE 6 is not permitted." Entering MODE 5 without SDM limits met implies that boron concentration in MODE 6 is not met. Under these conditions, a transition to MODE 5 should not be attempted until MODE 5 SDM limits are met. Inadvertent
- boron dilution events are precluded in MODE 6 via administrative controls
- [that close the dilution source valves], whereas dilution events are not
[ physically precluded] in MODE 5. Therefore, the transition from MODE 6 to i MODE 5 should not be allowed in the SDM initial condition for a MODE 5 dilution event is not met. INSERT 6A-4b TR 3.1-001 3.1-21 The ISTS SR 3.1.8.1 requirement to perfora a CHANNEL OPERATIONAL TEST (C0T) on the intermediate and power range NIS channels within 12 hours prior to initiating PHYSICS TESTS is revised to delete phrase "within 12 hours." COT testing is performed on these
- channels prior to reactor startup per LC0 3.3.1. This change is l consistent with traveler TSTF-108.
INSERT 6A-4c TR 3.1-003 l 3.1-22 The Completion Times for ITS 3.1.2, Required Actions A.1 and A.2 are increased from 72 hours to 7 days, consistent with traveler TSTF-142. I l l l l l [
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 3.1-003 APPLICABILITY: DC, CP, WC, CA REQUEST: Incorporate NRC-approved traveler TSTF-142 to increase the (CTS 3.1.1.5 AOT and] ITS 3.1.2 Required Actions A.1 and A.2 Completion Time from 72 hours to 7 days when the core reactivity balance is not within its limit. ATTACHED PAGES: Encl. 2 1-7 Encl. 3A 6 Enc!. 3B 4 Encl. 4 1, (new LS-24) Encl. 5A Traveler Status page,3.1-2 Encl. 58 B 3.1-11, B 3.1-12 Encl. 6A 4 Encl. 6B 3 l i
)
7 . l l REACTIVITY CONTROL SYSTEMS CORE REACTIVITY i l LIMITING CONDITION FOR OPERATION 3.1.1.5 The measured core reactivity shall be within 11% Ak/k of predicted values. APPLICABILITY: MODES 1 and 2 ACTION: 7 d*f ,7,, g With the measured core reactivity not within limits, within g 7;g,3,,, cog j
- a. reevaluate core design and safety analysis, and determine that the reactor core is acceptable for continued operation, and
- b. establish appropriate administrative operating restrictions and surveillance requirements, or
- c. be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.1.5.1 The overall core reactivity balance shall be compared to predicted values to den onstrate agreement within 11% Ak/k(once prior to entering MODE 1)
$$1MsM Cafter each refueling andlat least once per 31 Effective Full Power Days (EFPD) .
Plyg[] Qhereafter when bumup is >60 EFPD3 'nh - r :tr d:::::t et::r- ' ' "c% r:::':2- _ _ M M ?;::'- -- ' * *
- 15. The predicted reactivity values shall 3 3,Jg'
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. adjusted (normalized) to correspond to the actual core conditions prior to NO ex ing a fuel bumup of 60 EFPD after each fuelloading.
4.1.1.5.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k prior to initial operation above 5% RATED THERMAL POWER after each fuelloading, by consideration of the factors of Specification 4.1.1.1.1b, with the control banks at the maximum insertion limit of h%(] Specification 3.1.3.6. I 1 1 WOLF CREEK - UNIT 1 3/417 Amendment No. 89 f . t 1 1 Markup ofCTS3M.] S/15R7
CHANGE Nl#EER HSlE DESCRIPTION regarding the adequacy of the SDM with rods at their insertion limits is determined through compliance with ITS 3.1.2. which requires a reactivity balance prior to entering Mode 1 after each refueling. and ITS SR 3.1.6.1. which requires a verification of control bank position
~
within insertion 11mits within 4 hours prior to criticality. Therefore, the requirements of this SR would be performed by other specifications in the ITS [and by Physics Testing and the Reload Design Methodology). 05 05 LS 17 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 05 06 A " CTS SR 4.1.1.5.1 requires that the predicted reactivity values "shall" be adjusted (noma 11 zed) at 60-EFPD after ' refueling. ITS SR 3.1.2.1 states the normalization requirement as "may" be adjusted. This is to recognize that normalization is not necessary if predicted and measured core activity are within acceptable tolerance. The scheduling of the normalization of predicted and measured core reactivity continues to be required at 60 EFPD. Therefore, this change reflects clarification of
-, . existing intent and is considered administrative.
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LS-24 _ 06 01 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 07 01 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38).
. 1 07-02 -
Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). ; I 08 01 - Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 38). l 08 02 - Not applicable to WCGS. See Conversion Comparison Table ) (Enclosure 38). l 08 03 LS 19 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 09 01 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). i 10 01 - Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 3B). l WCGS-Description of Changes to CTS 3.1 6 5/15/97 L_ _____
INSERT'3A-6a TR 3.1-003 5-07 LS-24 The Allowed Outage Time (A0T) in the ACTION Statement of current TS LC0 3.1.1.5 is increased from 72 hours to 7 days. The proposed 7 day A0T is acceptable because of the conservatism used in designing the reactor core and in the performance of the safety analyses, as well as the low probability of a'DBA or anticipated
-transient approaching the core design limits occurring during the 7 day period. The proposed change relaxes the A0T associated with the measured core reactivity not being within 1000 pcm of the predicted value. The required ACTIONS call for a reevaluation of the core design and safety analysis, a determination of whether the reactor core is acceptable for continued operation, and the establishment of appropriate operating restrictions and SRs within 72 hours. The 72 hours allocated to perform these tasks is insufficient. Predicted versus measured reactivity anomaly evaluation is a complex proposition. Data would have to be gathered,-transmitted to the core design organization, evaluated by the fuel vendor, and implementation of appropriate controls
- would have to take place based on the data evaluation. Core design codes take time to set up for offnormal evaluations. RCS I boron samples would also have to be analyzed. Given the above l time-consuming activities, the proposed 7 day A0T is considered to be more realistic. More thorough troubleshooting and restoration l activities are possible with an extended A0T. This change is consistent with traveler TSTF-142.
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NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC) CONTENTS I. Organization ........................................................... 2 II. Description of NSHC Eval uations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 III. Generic No Significant Hazards Considerations A Administrative Changes.............................................. 5 R Relocated Techni cal Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 LG Less Restrictive (Moving Information Out of the Techni cal Speci fi cations) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 M More Restri cti ve Requi rements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 - IV. Specific No Significant Hazards Considerations LS LS 1.................................................................... 15 LS 2.................................................................... 17 LS 3.................................................................... 19 LS 4.................................................................... 21 LS 5.................................................................... 23 LS 6.................................................................... 25 LS 7.......................................................... ._ _ . LS.8..............................................,..........e...u..e..t. . . .. LS.9.................................................................... 35403 t-nl LS 10................................................................... 38 : LS 11..............................................................Not Used LS 12...................... ............................................ 40 LS 13................................................................... 43 LS 14................................................................... 45 LS 15............................................................. ..... 47 LS 16. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e ; LS 17........................................................Not applicable LS 18..................................................... .............. 49
- LS 19. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . Not appl i cabl e LS 20...........................................
i 8N. . . ..4....." 4.m. Not appl i cabl e LS 21. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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'LS 22.................................................................... 51 '
LS 23 . ._. . . . . . ._._. . . _ _. . . 1 ......_ --. _ .... .. . .
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' WCGS-NSHCs-CTS 3M.] 1 S/15/97
INSERT 4-b TR3.1-003] IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-24 10 CFR 50.92 EVALUATION i FOR TECHNICAL' CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS j l The Allowed Outage Time (A0T) in the ACTION Statement of current TS LCO 3.1.1.5 is increased from 72 hours to 7 days. The proposed 7 day A0T is acceptable because of the conservatism used in designing the reactor core and { in the performance of the safety analyses, as well as the low probability of a l DBA or anticipated transient approaching the core design limits occurring ] during the 7 day period. This change is consistent with traveler TSTF-142. 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 fac11ity involves no signifIcant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: i l 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 l
- 3. Involve a significant reduction in a margin of safety."
' The following evaluation is provided for the three categories of the I significant hazards consideration standards: i
- 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 proposed change relaxes the A0T associated with the measured core reactivity not being within 1000 pcm of the predicted value. The required ACTIONS call for a reevaluation of the core design and safety analysis, a determination of whether the reactor core is acceptable for continued operation, and the establishment of appropriate operating restrictions and SRs within 72 hours. The 72 hours allocated to perform these tasks is insufficient. Predicted versus measured reactivity anomaly evaluation is a complex proposition. Data would have to be gathered, transmitted to the core
i l IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATION NSHC LS-24 (continued) design organization, evaluated by the fuel vendor, and implementation of appropriate controls would have to take place based on the data evaluation. Core design codes take time to set up for offnormal evaluations. RCS boron samples would also have to be analyzed. Given the above time-consuming activities, the proposed 7 day A0T is considered to be more realistic. The proposed change in the A0T will not affect any of the analysis assumptions for , any of the accideats previously evaluated. The proposed change will not I affect the probability of any event initiators nor will the proposed change i 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 l 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 or different kind of accident from any accident previously evaluated?
l 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 change in A0T will not impact the normal method of plant operation. More thorough troubleshooting and restoration activities are possible with an extended A0T. 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, analysis l assumptions, methodologies, or credited equipment for any analyzed event. l 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. l There will be no impact on any margin of safety. 1 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS-24" 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.
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. l TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS Special Test Exceptions l 3.1.10-is retained and renumbered as 3.1.8, consistent with this l traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. I~'t 18 06F l , 1 TSTF-15, Rev.1 Incorporated NA NRC approved. TSTF-89 Incorporated 3.1-8 NRC approved. (TSTF-107, Ret. 1). Incorporated 3.1-6 . le 3.8 -15~ l l TSTF-108, N::i;;ger;;;d -NA-- Not NRC approved as-of Rev.1 incorporate { ].l 2.J __ _ N. TR 3.1-oal \ TSTF-110 Incorporated 3.1-10 - - 4 Rev. 2. M "PPY*
- W \*#
TSTF-136 3.1-9,3.1-15 Incorporated (1JRc hM-lTn.it- coc.j l TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoff date gg, u.o e [
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TSTF-142 N;; Ir. Ae_. yo_. ;;=ed T . : v i r i.. . e d e fte r incorporated _ __ 3. i -J2. gerage WPt.2greh C( Incorporated 3.1-7
'" % ..., 6 9 /@ 3 8- Ar I (WOG2Ydd Incorporated J 3.1-16
] TR. 3. 8 - 00(o]
1 5/15/97
Core Reactivity 3.1.32 h}$}E9;g 3.1 REACTIVITY CONTROL SYSTEMS 3.1.32 Core Reactivity I LCO 3.1.32 The measured core reactivity shall be within i 1% ak/k of ; predicted values. ! l 1 APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Heasured core reactivity not within limit. A.1 Re evaluate core design and safety analysis, h
,, 4 and determine that the reactor core is )
acceptable for continued operation.
'%' l A.2 Establish appropriate p .t y /
operating restrictions M_TK 3.t-0031 and SRs. B. Required Action and B.1 Be in H0DE 3. 6 hours ' associated Completion Time not met. I WCGS-Mark-up ofNUREG-1431-ITS 3.1 3.1 2 S/158 7
Core Reactivity B 3.1.32 BASES l LCO When mecsured core reactivity is within 1% ok/k of the predicted I (continued) v61ue at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These value's are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely, APPL;CABILITY The limits on core reactivity must be maintained during H0 DES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in H0 DES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing. In H0DE 6. fuel loading results in a continually changing core reactivity. Boron concentration requirements (LCO 3.9.1, " Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDH demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculation.s. Heasured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of _ is based on the low probability of a DBA occurring dur~ g t[is period, and allows sufficient time to l ra 3.l.6031 79.3* (continued) WCGS Mark-up ofNUREG-1431-Bases 3.1 8 3.1 11 S/158 7
Core Reactivity B 3.1.32 BASES assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. l Following evaluations of the core design and safety analysis, the ! cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate thet core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restrictions or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined. - 7 dy n.s.i-ons } The required Completion Time of is adequate for preparing whatever operating restr 1ons or surveillance that may be required to allow continued reactor operation. IL1 If the core reactivity cannot be restored to within the it ak/k limit, the plant must be brought to a H0DE in which the LCO does not appiy. To achieve this status, the plant must be brought to at least H0DE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by LD01.T;TRequf_ rect,,A.btfoMH S" 3.1.1.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.02.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 12 5/158 7
CHANGE NIABER JLISTIFICATION 3.1 17 Consistent with current TS LCO 3.1.3.2 and the wording of ITS 3.1.7 Conditions A and B. ITS 3.1.7 Condition C is clarified to state that the inoperable position indicators are inoperable DRPI's. 3.1 18 A MODE change restriction has been added to ITS 3.1.1 in the LCO Applicability, per the matrix discussed in CN 102 LS 1 of the 3.0 package. (See the LS 1 NSHC in the CTS Section 3/4.0. ITS Section 3.0 package). j q3,g.2; 3.1 19 Not used. 3.1 20 Consistent with current TS 3/4.10.3, " Physics Tests," ITS LCO 3.1.8 and its Condition C and SP. 3.1.8.2 are modified to refer to
" operating" RCS loops. Adopting the current TS is acceptable since valid T., measurements are not obtainable for a non-operating loop. .3.1-2.1 I bE.R.T -
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l l I WCGS-Differencesfrom NUREG-1431-ITS 3.1 4 5/15/97 l l t
INSERT 6A-4a 0 3.1-27 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. Ir. 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 LC0 3.0.4 were previously applicable in all MODES. ITS LC0 3.1.1 was modified by a Note stating: "While this LCO is not met, entry into MODE 5 from MODE 6 is not permitted." Entering MODE 5 without SLM limits met implies that boron concentration in MODE 6 is not met. Under these conditions, a transition to MODE 5 should not be attempted until MODE 5 SDM limits are met. Inadvertent boron dilution events are precluded in MODE 6 via administrative controls [that close the dilution source valves], whereas dilution events are not [ physically precluded] in MODE 5. Therefore, the transition from MODE 6 to MODE 5 should not be allowed in the SDM initial condition for a MODE 5 dilution event is not met. INSERT 6A-4b TR 3.1-001 3.1-21 The ISTS SR 3.1.8.1 requirement to perform a CHANNEL OPERATIONAL TEST (C0T) on the intermediate and power range NIS channels within 12 hours prior to initiating PHYSICS TESTS is revised to delete phrase "within 12 hours." COT testing is performed on these channels prior to reactor startup per LC0 3.3.1. This change is consistent with traveler TSTF-108. INSERT 6A-4c TR 3.1-003 3.1-22 The Completion Times for ITS 3.1.2, Required Actions A.1 and A.2 are increased from 72 hours to 7 days, consistent with traveler TSTF-142.
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 3.1-004 thru TR 3.1-006 APPLICABILITY: DC,CP, WC, CA REQUEST: Revise Traveler Status page to reflect NRC approval and latest revision number of travelers TSTF-14 Rev. 4, TSTF-110 Rev. 2, and TSTF-136. Change WOG-73 Rev.1 to TSTF-234 (still under NRC review). Remove traveler revision numbers everywhere except on the Traveler Status page. There are no changes involved to any CTS mark-ups, ITS mark-ups, DOCS, or JFDs. ATTACHED PAGES: Encl. 3A 1,3,4,10,13 Encl. 38 1,3,7,8 Encl. 4 25 Encl. 5A Traveler Status page Encl.6A 1,2,3 Encl. 6B 1,2,3 l l l i l l
DESCRIPTION OF CHANGES TO CURRENT TS Section 3/4.1 This enclosure contains a brief description / justification for each marked-up change to current Technical Specifications. The changes are identified by change numbers contained in Enclosure 2 (Mark up of the current Technical Specifications). In addition, the referenced No Significant Hazards Considerations (NSHCs) are contained l in Enclosure 4. Only technical changes are discussed: administrative changes (i.e.. l format, presentation, and editorial changes) made to conform to the improved l Technical-Specifications are not discussed. For Enclosures 3A. 3B, 4, 6A and 6B. ! . text in brackets "[ ]" indicates the information is plant specific and is not comon l to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE Nl#EER HSBC DESCRIPTION TA 3.1-ooG. } 01 01 LG In accordance with TSTF-9 this change would move the specified limit for Shu argin (SDM) from current TS to the COLR. This change occurs in several specifications including that for SDM and those specifications with ACTIONS that require verifying SDM within limits. SDM is a cycle specific parameter that is calculated based on an NRC approved methodology. Moving the SDM to the COLR will provide core design and operational flexibility that can be used for improved fuel management. 01 02 M s modific on redefines t applicab lih catio o inclu ith kg Mode 1.0" n [thes add rrent on to es 3. cifica n
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(an S Spec cation . 6) defi the S down gin j plicabi ' y requi nts for de 1 a de 2 th q a 1. The pro ed change uld be e res 1cti t ( wo d repre only a s 1 change t curr SDM (applicab'ityrequir ts. 145ERT 34-1 o [ Q3.1 1 01-03 LS 1 The Action Statement would be modified to reflect that the requirement to initiate boration at a specified rate with fluid at a specified boron conGntration is generalized to simply require boration. As described in the ITS Bases, the required flow rate and boron concentration should be , selected depending on plant conditions and available j l equipment. The ITS Bases allow the operator to use the l "best source available for the plant conditions. This is an example of maintaining the overall safety requirement in TS but removing procedural details from the TS allowing I the plant operator the ability to select the appropriate procedure and equipment for the existing plant condition. WCGS-Description of Changes to CTS 3.1 1 5/158 7 l
CHANGE NUMBER gilE DESCRIPTION 02 01 A In the conversion process this LC0 will be combined with the SDM LCO applicable for T.,, > 200*F. in accordance with -- traveler TSTF 136. Traveler TSTF 9.(jefVfJrelocated !TM 'M values for SDM to the COLR which removed tne only difference between ISTS LCO 3.1.1 and ISTS LC0 3.1.2 Differences above and below 200*F will be addressed in the COLR. 03 01 A The footnote referring to Special Test Exceptions would be deleted. This is acceptable because the requirements for Special Test Exceptions are provided _in separate LCOs. Therefore, a separate reference in the footnote is redundant. 03 02' LS 3 Action Statement A.1 would be revised to require achieving Mode 2 with k,y < 1.0 instead of achieving HOT STANDBY if the BOL HTC limit is exceeded and revised rod withdrawal limits cannot be established. This change corrects the discrepancy between the BOL Applicability and the Action, while ensuring that the plant is taken to a condition in which the LCO is not applicable. Revising the current TS, albeit to correct an inconsistency, represents a relaxation in Action Statement a.1. 03 03 A The statement that administrative withdrawal limits required to meet Action Statement a.1 are in addition to insertion limits of another specification would be removed. This change is an administrative change because ! the statement is redundant to the requirement of current ' TS 3.1.3.6 and therefore can be deleted. 03 04 LG The requirement of current TS Action Statement a,2. which provides the details of how to verify that MTC has been restored to within limits (i.e., calculation) for the all rods withdrawn condition prior to exiting Action Statement a.1 is addressed in the ITS 3.1.3 Bases. 03 05 TR 2 The requirement to submit a Special Report to the NRC would be deleted. This is in conformance with the ISTS. 03 06 LS 4 This change would incorporate a Note from ITS 3.1.3 allowing suspension of MTC testing near the end of the I cycle when further significant changes to the MTC would not occur and result in exceeding the EOL limit. This represents a relaxation in performing the surveillance requirement. WCGS-Description of Changes to CTS 3.1 3 SMSM7
CHANGE J NUMBER HSE DESCRIPTION l 03 07 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 04 01 LS 5 This proposed change would make two changes to the Action Statement. First, it would alter the Action Statement shutdown requirement time limit from a combination of 15 minutes to restore T.,, to within limits followed by 15 minutes to be in Mode 3, if T.,, could not be restored, to a single 30 minute limit to exit the Applicability if T , were not within its limit. Second, the Action Statement would be revised to require achieving Mode 2 with k,y < 1.0 instead of achieving HOT STANDBY if the LCO is not met (refer to TSTF 26). Regarding the first change, both the current requirement and the ITS requirement are essentially equivalent in that the plant is now required to shutdown and exit the Applicability of the specification within 30 minutes after discovering that a parameter is not within its limits, if the parameter is not restored to within its limits in that 30-minute time period. If the LC0 can be satisfied at any time during the 30 minute time frame, the plant shutdown can be terminated. Regarding the second change, it represents a relaxation in current Action Statement requirements for i plant shutdown, consistent with exiting the LCO's l Applicability. 04 02 LS 6 The proposed change would revise the Surveillance Requirement for verifying that Reactor Coolant System (RCS) temperature (T,,,) is within limits by changing the j Frequency to once per 12 hours in accordance with traveler l TSTF-27M The current frequency requirements arefrus-ooq within 15 minutes prior to achieving reactor criticality, which is redundant and unnecessary since T.,, must be within its limit prior to entering the LCO Applicability, and at least once per 30 minutes when the reactor is critical and the (T,,,-T,,) Deviation Alarm is not reset. The RCS temperatur.e is maintained within limit: (1) to l assure that the Moderator Temperature Coefficient is l within the limits assumed in the accident analysis: (2) to l assure that the neutron detectors are not adversely l affected: (3) to assure that the RCS and pressurizer response to thermal hydraulic transients is as predicted: and (4) to assure that the reactor vessel temperature is above the mil ductility transitian reference temperature. The plant design incorporates the monitoring of T,,, and provides an alarm, the (T,,, T,,) Deviation Alarm, as T,,, approaches its limit. This alarm condition requires a WCGS-Description ofChanges to CTS 3.1 4 5/158 7
CHANGE Nt#EER HSE DESCRIPTION This more restrictive change clarifies the appropriate ACTIONS to be taken for all causes of inoperability, consistent with traveler TSTF-107. 12 15 A Rod misalignment is determined based on a comparison between the rod's DRPI and its group step counter demand position,' not by a rod to rod position verification. This change is administrative in nature in that there is no effect on the manner in which the operating staff would determine whether a misalignment event had occurred. This change is consistent with NUREG 1431, Rev.1. 12 16 LG Several surveillance (e.g., rod position deviation monitor and rod insertion limit monitor in this section) contain actions in the form of increased surveillance frequency to be performed in the event of inoperable alarms. These actions are moved from the TS to licensee controlled doctments since the alarms do not themselves directly relate to the limits. This detail is not required to be in the TS to provide adequate protection of the public health and safety. Therefore, moving this detail is acceptable and is consistent with traveler TSTF-110 TA3._co4.] 12 17 - Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 38). l 12 18 LG The technical contents of the Action Statement which allow l continued power operation with a misa11gned rod are moved to the Bases for ITS LCO 3.1.4. Action B.1. i 12 19 LS 18 This change revises current TS Surveillance Requirement [4.1.3.1.2] frequency 31 days to 92 days. This surveillance requirement exercises each rod cluster control assembly at least ten steps to demonstrate the . ability of the rods to be tripped. This change is based I on the recommendations of Generic Letter 93 05 and is i compatible with plant operating experience. This change i is also consistent with NUREG 1431. Rev.1. 1 12 20 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 12 21 - Not used. 12 22 M This change, in accordance with NUREG 1431. Rev. 1. provides a new ACTION in the event the allowed outage times are not met for the rod misalignment actions. Prior WCGS-Description of Changes to CTS 3.1 10 5/13/97
CHANGE NUMBER NSE DESCR1PTION 13 08 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 13-09 LS 23 Current TS Actions b.1.b and b.1.c of LCO 3.1.3.2 are deleted. SDM is ensured in MODES 1 ana 2 by rod position. Multiple inoperable DRPIs will have no impact on SDH in j MODES 1 and 2 if the control rod position are verified by i alternate means and rod motion is limited consistent with the accident analysis. Deletion of these requirements is i consistent with traveler (OiHJ3r3er'J j ne. 3,s.aer.) i mar 24Q q A i-n 1 t-CfsVF -234D 14 01 - _ Not applicable to WCGS. See conversion Comparison Table (Enclosure 38). 15 01 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 15 02 - Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 16 01 LS-14 This TS would be revised to apply to shutdown " banks" instead of shutdown " rods": this is consistent with NUREG-1431. Rev. 1. The current Action Statement permits one ' rod to be inserted beyond the limits: the proposed ITS i CONDITION A would allow one or more banks to be inserted beyond the limit. 16 02 M The proposed changes to the Action Statement would require l that the shutdown banks be aligned within limits and that SDM be verified or restored. The new Action Statement would extend the time to achieve alignment from 1 to 2 hours as justified in the Bases for ITS 3.1.5. The new Action Statement would establish a Completion Time of 1 hour for verifying and restoring SDM. In the proposed Action Statement, both the realignment and the SDM verification would be required. The current Action Statement provides a 1 hour limit to achieve realignment and effectively applies a 2 hour Completion Time to SDM verification and restoration Nhich would be performed under the TS for rod group ahCnment limits). In the current Action Statement, either the realignment or the SDM verification are required. The current Action i Statement could, in some circumstance, allow continued POWER OPERATION with a shutdown rod out of alignment because it was written to apply to individual rods and refers to the rod group alignment specification. The new action statement, which applies to shutdown banks, would not permit operation with a shutdown bank outside its WCGS-Description of Changes to CTS 3.1 13 SAS/97
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IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 6 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The proposed change would revise the Surveillance Requirement for verifying that Reactor Coolant System (RCS) temperature (T,) is within limits by changing the- p_ooq Frequency to once per 12 hours in accordance with industry traveler TSTF 27(# The original frequency requirements were within 15 minutes prior to achieving reactor criticality, which is redundant and unnecessary since the To must be within its limit ' l prior to entering the LC0 Applicability, and at least once per 30 minutes when the reactor is critical and the (T -Tm) Deviation Alarm is not reset. The RCS temperature is maintained within limit: (1) to assure that the Moderator Temperature Coefficient is within the limits assumed in the acident analyses (2) to assure that the neutron detectors are not adversely affected by neutron attenuation caused by low coolant temperature, (3) to assure that the reactor coolant system and pressurizer l response to thermal-hydraulic transients is as predicted, and (4) to assure that the l reactor vessel temperature is above the nil ductility transition reference temperature, f The plant design incorporates monitoring of T and provides an alarm, the (T T,) Deviation Alarm, as T m approaches its limit. This alarm condition requires a response by the operating staff. Therefore, at any time that T is approaching its limiting value, the plant operators would receive an alarm and initiate corrective l action. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in i accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:
"The Connission may make a final determination. pursuant to the procedures in 50.91. that a proposed amendnent to an operating license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no !
significant hazards consideration. if operation of the facility in accordance l with the proposed amendnent 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: WCGS-NSHCs-CTS 3N.] 25 5/1587 t
INDUSTRY TRAVELERS APPLICABLE TO SECTION 3.1 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 Incorporated 3.1-1 NRC approved. TSTF-12, Rev.1 Incorporated 3.1-15 NRC approved. ITS l Special Test Exceptions 3.1.10 is retained and renumbered as 3.1.8, ! consistent with this I traveler and TSTF-136. TSTF-13, Rev.1 Incorporated 3.1-4 NRC approved. TSTF-14, Revh) Incorporated 3.1-13 NRC approved. I ne ss.m l TSTF-15, Rev.1 Incorporated NA NRC approved. l TSTF-89 Incorporated 3.1-8 NRC approved. D Incorporated 3.1-6 . [42.1-1s_] TSTF-108,3 N::!E:gxi:;2- -NA- Net NRC approv'ed as.ef R** 4 _ _ _ incorPorpeg _ y _2]. _ ___ , N. Ta s.i.onTI TSTF-110 Incorporated 3.1-10 --> Rev.h hEqr "T R '3.1-oo4 TSTF-136 Incorporated 3.1-9,3.1-15 (N E M-\T a.s.t- coc.) TSTF-141 Not incorporated NA Disagree with change; traveler issued after cutoff date TSTF-142 B';; : ..I y .. k 2 -NA-- T.Tv;;u :.. ed ef:cr#* *** I incorporer4A 3.1- 22 _ eger g;e WPt.agrek ( Incorporated 3.1-7 (WOG2dR Incorporated i 3.1-16 ? ".Z. . y "9/@ s 8- ai
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DIFFERENCES FROM NUREG-1431 Section 3.1 l This enclosure contains a brief discussion / justification for each marked-up technical change to NUREG 1431. Revision 1, to make them plant specific or to incorporate generic changes resulting from the Industry /NRC generic change process. The change numbers are referenced directly from the NUREG 1431 mark ups. For Enclosures 3A, 3B, 4. 6A, and 6B. text in brackets "[ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUMBER JUSTIFICATION TR. 3.1-obro l 3.1 1 In accordance with TSTF 9, #,4 this change would relocate the specified limit for Shutdown rgin (SDM) from the ISTS to the COLR. This change occurs in several specifications including the Specification for SOM and those specifications with ACTIONS that require verifying SDM within limits. 3.1 2 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 6B). Q3.1-23 3.1-3 Wolf Cree ITS L 3.1. Requ1r Acti C.1 i rev eda' rom ~ ! in MOD . Be i E ith < 1.0 " T icabi y of 3. is E1a 2 wit ,, 2 .. a ition, he ses scuss n indi ed t equir Act' ns' ntent 's to pl the it in Ew the is ta icab : this uld MODE 2 ith K,, 1.0. his nge simffartothe (chan propos in tr eler F 26. M tlwl. 3.1 4 SR 3.1.4.2 of NUREG-1431. Rev. I would be deleted. In accordance l l with TSTF 13.9WD the intent of this SR is only to determineM2.1 oor ) the next frequency for SR 3.1.4.3. Performance of SR 3.1.4.2 is not necessary to assure that the LC0 is met: SR 3.1.4.3 fulfills that purpose. Therefore, SR 3.1.4.2 may be deleted. In addition, the Note in the Frequency column of SR 3.1.4.2 would be moved to become Note 1 in the Surveillance column of SR 3.1.4.3. This is for clarification purposes. As discussed in CN 3.19, section re ntsnbering results in SR 3.1.4.3 of NUREG 1431. Rev.1 ! becoming ITS SR 3.1.3.2. 3.1-5 Per current TS [3.1.3.1], the words "with all" have been removed from ITS LC0 3.1.4. This is a clarification that ensures the proper interpretation of the LCO. The change makes it clear that only one channel of DRPI is necessary to meet the alignment accuracy requirement of the LCO. k'ith the word "all" in the WCGS-Differencesfrom NUREG-1431-ITS 3.1 1 S/1S/97
CHANGE NUMER JUSTIFICATION statement. it may be possible for those unfamiliar with the DRPI design to interpret the LCO as applying to all channels of DRPI. 3.1-6 ITS LCO 3.1.4 would be split into two separate statements to clarify that the alignment limit is separate from OPERABILITY of the control rod. The COW ITION A wording is broadened from "untrippable" to " inoperable" to ensure the CONDITION encompasses all causes of inoperability. Previous wording was ambiguous for , rods that, for instance, had slow drop times but were still j trippable. These slow Ns are inoperable rods, and the change clarifies the appropriate ACTIONS. The Bases are changed to reflect the changes to the LCO and CONDIT. ION A. These changes are based on traveler TSTF 107. 3.1 7 This change to the ISTS would incorporate, into LC0 3.1.7. an Action Statement that was previously approved as part of the Callaway and Wolf Creek licensing basis, as revised in Enclosure
- 2. The Action Statement would permit continued POWER OPERATION for up to 24 hours with more than one Digital Rod Position Indicator per rod group inoperable. The Action Statement specifies additional required actions beyond those applicable to the condition of one DRPI channel per group inoperable. The Bases for this change also would be incorporated into the Bases for the )lant ITS. These changes are consistent with traveler FMtef._%. The note under the ACTIONS is changed to be consistent with the new Required Actions.
3.1 8 The Frequency for ITS SR 3.1.7.1 for comparing DRPI and group demand position would be changed from 18 Months to "Once prior to i criticality after each removal of the reactor vessel head." This ! change makes it clear that the surveillance must be performed j each time the head is removed and that it is not tied to an ) absolute time interval. This change is based on traveler TSTF- j
- 89. l 3.1 9 This change would eliminate ISTS 3.1.2 because the SDM requirements for MODE 5 have been incorporated into Specification ,
3.1.1 in accordance with traveler TSTF 136. Traveler TSTF 9. ; Rev.1, relocated values for SDM to the COLR which removed the only difference between ISTS LCO 3.1.1 and ISTS LCO 3.1.2. Differences above and below 200'F will be addressed in the COLR. 1
- Subsequent sections have been renumbered.
4 3.1-10 Several surveillance (e.g., rod position deviation monitor and rod insertion limit monitor in this section) contain actions in the form of increased surveillance frequency to be performed in the event of inoperable alarms. These actions are moved from the WCGS-Dufferencesfrom NUREG-1431-ITS 3.1 2 5/1587
CHANGE EtER JUSTIFICATION TS to licensee controlled documents since the alarms do not themselves directly relate to the limits. This detail is not required to be in the TS to provide adequate protection of the public health and safety. Therefore, mving of this ail is acceptable and is consistent with traveler TSTF 110 Tgu-ea4 ) 3.1-11 Not used. j 3.1 12 The Required Actions for inoperable DRPI in ITS. 3.1.7 are revised per the current licensing basis to note that the use of movable incore detectors for rod position verification is an indirect , assessment at best. The position of some rods can not be ! ascertained by this method. 3.1 13 This change adds an LCO requirement and SR to MODE 2 Physics Tests Exceptions 3.1.8 to verify that themal power is less than or equal to 5 percent RTP. The LCO requirement and SR were added to verify that Thermal Power is within the defined power level for MODE 2 during the performance of Physics Tests, since there is an Action that addresses Thermal Power not within limit yet there was no corresponding LC0 or surveillance requirement. The ! Surveillance Frequency of 1 hour is retained _from the current TS. j This change is based on traveler TSTF-16 ra as-oos] i 3.1-14 Not used. 3.1 15 Consistent with TSTF-12. Revision 1. LCOs 3.1.9 and 3.1.11 are deleted. The physics tests contained in LCO 3.1.9 were only contained in some plant initial plant startup testing programs. The physics test can be deleted since these physics tests are never performed during post refueling outages. The physics test that LCO 3.1.11 required was the Rod Worth Measurement in the N 1 condition. The use of other rod worth measurement techniques will maintain the shutdown margin during the entire measurement process and still provide the necessary physics data verification. Since the N 1 measurement technique is no longer used, the SDM test exception can be deleted. This change and traveler TSTF 136 renumbers ITS 3.1.10 to 3.1.8. 3.1 16 This change adds the requirement to perform SR 3.2.1.2 in addition to SR 3.2.1.1 during performance of ITS 3.1.4 Required Action B.2.4.' The intent of Required Action B.2.4 is to verify the Fa(Z) is within its limit. Fa(Z) is approximated by F/(Z) (which is obtained via SR 3.2.1.1) and Fj(Z) (which is obtained i via SR 3.2.1.2). Thus both Ff(Z) and F/(Z) must be established to verify Fa(Z). This change is consistent with traveler WO3105 (later) . 3 S/158 7 WCGS-Differencesfrom NUREG-1431-ITS 3.1 1
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.1-ED APPLICABILITY: WC REQUEST: 1) Bases 3.1.4, Background, delete the word " control" consistent with deletion in ITS 3.1.7 Bases Background. ,
- 2) Bases 3.1.4, LCO, retain the STS word " OPERABILITY" instead of "LCO" that l was added. I
- 3) Bases 3.1.3, SR 3.1.3.2 delete "not" in the first sentence. The ITS SR wording j is a not required to be performed surveillance and the Bases wording was '
developed to indicate when the surveillance is performed. ATTACHED PAGES: Encl. 5B B 3.1-20, B 3.1-22, B 31-24 1 i
T MTC B 3.1.43 BASES SURVEILLANCE SR 3.1.43.2 =d S" 3.1.4.3 (continued) REQUIREMENTS 1., The:SRlishrequiredf,'to_beperformedioncelea.gh; cycle withinZeffectivelfull t powerrdaysILEFPDsy'after~ reaching tKaruivalent or=an:equilibrM_RIEallirodslouti(ARO) leoncentrationlof~300lppe;; s.n.u-s
- g. 2. If the 300 ppm Surveillance limit is exceeded, it is j possible that the E0C limit on MTC could be reached before the planned E0C. Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is l sufficient to avoid exceeding the E0C limit. j 03.1.fr-11
} 3 ~.' The Surveillance limit for RTP boron concentration of )
l 60 ppm is conservative. If the measured MTC at 60 ppm is j lessnegatjveser;p;;itivethanthe60ppmSurveillance limit, the E0C limit will not be exceeded because of the j 1 gradual manner in which HTC changes with core burnup. 1 REFERENCES 1. 10 CFR 50. Appendix A. GDC 11. j
- 2. FSAR USAR, Chapter.f15}.
- 3. 'iCA" ^273 N" A. "'.Jc;tingeu;c "al;;d Safety 0;;1ustian
";ttedelsi;y," July 1^05. NSAGj007.ZReloal;SafetV Evaluation: Methodology for the1 Wolf; Creek Generating StatigZ
- 4. ISA" Ctepter [15].
I WCGS-Mark-up ofNUREG-1431-Bases 3.1 ' B 3.1 20 5/1587
Rod Group Alignment 1.iaits l B 3.1.54 l BASES l l BACKGROUM) GroupsenrithinXbank are moved in a staggered fashion, but (continued) always within one step of each other. All r,it; Mv; few l .;;rtr;l kri; ;r.d ;t 1;;;t ta; ; btd;r. E ri;. ATDcontroYbanks CERillMNJEneCh. shutdown.3lanksmgi!dlariBILngdK]Mo gnRKanEtha/Mhaining;three ' shutdown 2anks"(CiW an5 %) conteveargreep; The shutdown banks are maintained either in the ' fully inserted or j fully withdrawn position. The control banks are moved in an l overlap pattern, using the following withdrawal sequence: When ! control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maxists withdrawal, and control bank B continues to move out. When control bank B reaches a
- l. predetermined height, control bank C begins to move out with j control bank B. This sequence continues until control banks A.
l B, and C are at the fully withdrawn position, and control bank D 4 l 1s approximately halfway withdrawn. The insertion sequence is ! the opposite of the withdrawal sequence. The control rods are ; arranged in a radially symmetric pattern, so that control bank l motion does not introduce radial asymmetries in the core power distributions. The axial position of shutdown rods and control rods is indicated I by two separate and independent systems, which are the Bank Demand Position Indication System (connonly called groep step counters) and the Digital Rod Position Indication (DRPI) System. The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at i the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered l highly precise (* 1 step or i % inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. The DRPI System provides a highly accurate indication of actual Occ3.1-EDQ~ui rod position, but at a lower precision than the step t counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. with ; ca.ter t
;;..:.;r di;t;ra; cf 0.75 irch;;, which i; ;ix ;tep;. To increase (continued)
WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 22 S/15M7
Rod Group Alignment Limits B 3.1.54 BASES APPLICABLE bounds the situation when a rod is misaligned from its group by SAFETY ANALYSES 12 steps. (continued) Another type of misalignment occurs if one RCCA fails to insert l upon a reactor trip and remains stuck fully withdrawn. This ! condition is assumed in the evaluation to determine that the i required SDH is met with the maximum worth RCCA also fully l withdrawn (Ref. 35). l The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERHAL POWER will be adjusted so that excessive local linear heat rates (LtRs) will not occur, and that the requirements on SDN and ejected rod j worth are preserved. Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (Fo(Z)) ! and the nuclear enthalpy hot channel factor (FIw) are ! verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misa11gned, the assumptions that are used to determine the rod insertion limits. AFD limits, and quadrant power tilt limits are i not preserved. Therefore, the limits may not preserve the design peaking factors, and Fa(Z) and Flu must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) l contains more complete discussions of the relation of Fa(Z) and Flu to the operating limits. l Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDN. which are initial
- conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of the =0 "clicy St;tsat 10;CFR,50;36(c)(2)(11)_.
LCO The limits on shutdown or control rod alignments ensure that the asstmotions in the safety analysis will remain valid. The N8 l'# l nut, requirements on*0PERABILITY ensure that upon reactor trip, the w d d d a ga. reactivity will be available and will be inserted. The IP t- A LO"C"'"!LI @ requirements M ensure that the RCCAs and banks ( maintain Ine correct power distribution and rod alignment. Tlg ( rod l.0PERABILITLrequirement11 s7satisfiediprovjdedithelfod' 1 s trippable and meets the rod drop time requirements of SR 3.1.4.3. l are sepwAt bm +h sh 3 n= q;s,i is ) requeamenh pWch. (continued) WCGS-Mark-up ofNUREG-1431-Bases 3.1 B 3.1 24 S/15/97 l
Attachment 2 to WO 98-0078 Page 1 of 4 t JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS h CTS 3/4.2 - POWER DISTRIBUTION SYSTEMS ITS 3.2 - POWER DISTRIBUTION SYSTEMS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES l 1 1
Page 2 of 4 INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER 3.2.G-1 DC, CP, WC, CA YES 3.2-1 CP NA 3.2-2 DC NA 3.2-3 DC, CP, WC, CA YES 3.2-4 DC, CP, WC, CA YES 3.2-5 CA NA 3.2-6 DC, CP, WC, CA YES 3.2-7 WC, CA YES 3.2-8 WC YES 3.2-9 WC YES 3.2-10 DC, CP, WC, CA YES CA 3.2-001 DC, WC, CA YES CA 3.2-002 DC, WC, CA YES CP 3.2-ED CP NA CP 3.2-001 DC, CP, WC, CA YES DC ALL-005 (3.2 changes only) DC NA DC 3.2-ED DC NA DC 3.2-001 DC NA TR 3.2-004 DC, CP, WC, CA YES WC 3.2-001 CP, WC YES 1 l _. - - - . _ _ )
Att; chm:nt 2 to WO 98-0078 P ge 3 of 4 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:
- 1. Each licensee is submitting a separate response for each section.
l
- 2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that l 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 l for brackets, where applicable (using the same methodology used in enclosures 3A,3B, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as PLANT SPECIFIC 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 i the clearest possible indicction 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 I to reflect the changes being made by one or more of the other licensees. These changes are not included in the additional information for the licensee to which the i change does not apply, as the changes are only for consistency, do not technically affect the request for that licensee, and are being provided in the additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment i request supplement to be provided later.
l l I- __ _ _
Att: chm:nt 2 to WO 98-0078 Pagm 4 of 4 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued)
- 8. The item numbers are formatted as follows: [ Source][lTS Section)-[nnn)
Source = Q - NRC Question CA- AmerenUE DC-PG&E WC - WCNOC CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL" is used for the section number. nnn = a three digit sequential number or ED (ED indicates editorial correction with no impact on meaning) l I i l I
_ _ _ _ _ - - - - _ _ - = _ - - _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ - _____-_____- __- ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2 G-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.X Bases General There have been a number of instances that the specific changes to the STS Bases are not properly identified with redline or strikeout marks. Comment: Perform an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal. FLOG RESPONSE: The submitted ITS Bases markups for Section 3.1 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorialin nature and would not have affected the review. Examples of editorial changes are:
- 1) Capitalizing a letter with only a " redline" but not striking out the lower case letter that it replaced.
- 2) Changing a verb from singular to plural by adding an "s" without
" redlining" the "s".
- 3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g., SR3.6.6A.7).
- 4) Changing a bracketed reference (in the reference section) with only a " redline" for the new reference but failing to include the strike-out of the old reference.
- 5) in some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.
- 6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.
- 7) Additional text not contained in the STS Bases was added to the ITS Bases by the lead FLOG member during the development of the submittal. Once it was determined to not be applicable, the text was then struck-out and remains in the ITS Bases mark-up.
Differences of the above editorial nature will not be provided as attachments to this response. The pages requiring changes that are more than editorial and are not consistent with the markup methodology are attached. ATTACHED PAGES: l l Encl.5B B 3.2-3, B 3.2-7, STS B 3.2-20, B 3.2-19, STS B 3.2-42, B 3.2-26 1 u___________...._._____.
Fa(Z) (Fa Methodology) B 3.2.1 BASES LCO The Heat Flux Hot Channel Factor, Fa(Z), shall be limited by the { following relationships: ' rjz) s em x(z) for P > 0.5 P F/z) s 0.5 K(z) for P s 0.5 where: CFQ% Fa" is the Fa(Z) limit at RTP provided in the COLR, K(Z) is the normalized Fo(Z) as a function of core height provided in the COLR, and THERMAL POWER p, RTP For thi; f;;ility Tthe actual values of CFQ and K(Z) are given in the ! COLR.:. { X(Z' i; e farctier, th;t leek; lik; th; era previded in i Figer; S 0.2.10 1. l For Relaxed Axial Offset Control operation, Fa(Z) is approximated by Q(Z) and Pd(Z). Thus, both Q(Z) and Pd(Z) must meet the preceding limits on Fa(Z). An Q(Z) evaluation requires obtaining an incore flux map in MODE 1. From the incore flux map results we obtain the measured value (Pd(Z)) of Fa(Z). Then, red h*
$(Z) = Pd(Z) [1.0015]bl.03 f (1.0 =3 qg 3.2.6-1 (
wher; [1.0015] IIO3 is a factor that accounts for fuel manufacturing tolerances and105 isla factor,;that~accountslikr t flux map measurement l uncertainty. Q(Z) is an excellent approximation for Fa(Z) when the reactor is at the steady state power at which the incore flux map was taken. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.2 B 3.2 3 5/I58 7 l ) L
Fe(Z) (Fa Methodology) B 3.2.1 BASES
$3 Qui % rib camAiW steamicAA a*j SURVEILLANCE of Q(Z) and F#(Z) follcwing a power increase of more than 10%, ensures REQUIREMENTS that they are verifiedTsERuMrflyset1#aSWP3RTP (or any other (continued) level for extended operation) Cia acmevan Equmbriin: conditions 1are achfeMMienTthelcore.;isisuff_iSijuOistablMJ4is[M .
op gatsjpg M t A1tofperfttre.J1Kasppi!!g; In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of Q(Z) and F#(Z). The Frequency condition is not intended to require v' verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which Fa was last measured. SR 3.2.1.1 , Verification that $Z) is within its specified limits involves increasing F#(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain $Z). Specifically, F#(Z) is the measured value of Fo(Z) obtained.from incore flux map results and 41.0815 ef. 4). Q(Z)isthencomparedtoits
$Z) = F#($(inu.
specified lii p .3. g. e . } The limit with which $Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR. 4 Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the Q(Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased. If THERMAL POWER has been increased by a 10% RTP since the last determination of Q(Z), another evaluation of this factor is required C312]%q after achieving equilibrium conditions (@7J' TSP) p Tii W W p n,..A t thi: hi # a i,ca - 1: x1 (to ensure that Fa'-(Z) values ~ are being reduced sufficiently with power i_ncrease to stay within the LCO limits). L gwlg,_ Q-), { l The Frequency of 31 EFPD is adequate to monitor the change of power i distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS). (continued) WCGS-Mark-up ofNUREG-1431 Bases 3.2 B 3.2-7 S/15/97
Fa(Z) (Fo Methodology) B 3.2.1B 1.2 DO NOT OPERATE IN THIS AREA (6.0,1.0) k N ( .8, 0.9 4 ) 0.8 '
\
(12.0,0.65) y 0.6 r 0.4 f
/
THI FIGURE FOR ILL TRATION ONLY. , 0.2 O NOT USE FOR OPERATION FT. 0 / 2 4 6 8 t 12 (*) % 16.6 33.3 50.0 66.7 83. 100 CORE HEIGHT
'F core height of 12 feet 9, }ep, A Ceepe. %)
Figure B 3.2.181 (page 1 of 1) K(Z) Normalized Fo(Z) as a Function of Core Height G 3.2.&-1 J WOG STS B 3.2-20 Rev 1, 04/07/95
AFD (RAOC Methodology) B 3.2.3 BASES
-- / _r #- h_ -
.s BACKGROUND rovides an alars messas; fast;t;1y if_the_AFD for two or moref 4
(continued) OPERABLELexcore f:harinels]sfoutside;;1ts specified.ljnits. Ns24 } ) APPLICABLE The AFD is a measure of the axial power distribution skewing SAFETY ANALYSES to either the top or bottom half of the core. The AFD is sensitive to many core related parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration. The allowed range of the AFD is used in the nuclear design process to confirs that operation within these limits produces core peaking l factors and axial power distributions that meet safety analysis 1 l requirements. The RAOC methodology (Ref. 21) establishes a xenon distribution library with tentatively wide AFD limits.- Cm di ;..;isi si;l Axtal power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the' safety analysis requirements. The limits on the AFD ensure that the Heat Flux Hot Channel Factor (Fa(Z)) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of ConditiorM3 g, B-{H, or +-IV events. This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition +-H event is the LOCA. The most important Condition S-E event is the loss of flow accident. The most important Condition E-II events are uncontrolled bank withdrawal and boration or dilution accidents. Condition E-II, accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower aT and Overtemperature AT trip setpoints. The limits on the AFD satisfy Criterion 2 of th: OC I;1 icy Ct; tac.; 10;CFR;50Mc)J2) lid. (continued) WCGS-Mark-up ofNUREG-H31 - Bases 3.2 8 3.2 19 $/15M7 1
r ,. AFD (RAOC Methodology) B 3.2.3B N /
\ /
x (-15,100) (6,100) UNACCEPTABLE \ \ UNACCEPTABLE OPERATION OPERATION 80 N
\
O Q _ ACCEP BLE OP TION \ - 1 ss-(-31,50) ( 50) 40 l
/
20 Till IGURE IS FOR IL TRATION ONLY. NOT USE FOR PERATION. \ l
-10 10 30 0
-50 -30 40
' -40 -20 0 20 AXIAL FLUX DIFFERENCE (%) hteked ' (emyte.on% ) l Figure B 3.2.3B-1 (page 1 of 1) IQ 3.2.cs-1 i AXIAL FLUX DIFFERENCE Acceptable Operation Limits I as a Function of RATED THERMAL POWER B 3.2-42 Rev 1, 04/07/95 WOG STS o - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _
QPTR B 3.2.4 BASES realme. ACTIONS cp 3. 2. & - 1 (continued) A reduction,of the Power Range Neutron Flux-High;tr,ip setpoirttp~)y 4 forceach:1r by which100R _egeodu;oonis_a conserymtsammetLion fot;protectjonJgainst the,consequencesiofiseyeteittansieggLggb potentiallylunanalyzed poWetdistrjbutions] ~fetformance:M RequiredLAction;results inLear11er; trip;setpoint:teductioLggg.glould be; required. pursuant to_the Required 3ctionsiof_.then anGdB3 specifications;:The completioctmeE72 hours ^after_eacturm deteminationjs; sufficient considettagithelsmalElikeljhogGD severe 1 transient in this time period 1and theLpreceding proigt reduction in fer as.z; THERMAL 4QQ .s.2-c POWER l_ in accordanceiwith Required A
)qa.2.0-\
U Although FL, and Fo(Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core , power distribution. This re evaluation is required to ensure that, before increasing THERMAL POWER to above the limits of Required Actions A.1 and'A.4, the react core conditions are consistent with the assumptions in the safety nalyses. _ __ a.A Paaer b5
- Aha- m.z-q A-6A.6 Flu-Hi$ pp setgeid -
If the QPTR bs en;;d;d remains,above the 1.02 limit and a
-_ __ re evaluation of the safety analysis is completed and shows that rM QFrr2 safety requirements are met, the ex_ core _dete_ctors arq re;;libreted t;
.te, A e ;;re = normal 12editcNtMdias4tikM prior to thA 1.k.}.J increasing THERMAL POWER to above the limit of Required Action A.1.
l The; process of; normalization isfaccomplished by measuringLcurrentsifor i each detecfor_during fluxLaapping and j l (continued) i l WCGS-Mark-up ofNUREG-1431. Bases 3.2 B 3.2 26 S/15A7 l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-3 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.1 Heat Flux Hot Channel Factor CTS 3/4.2.2 Heat Flux Hot Channel Factor (All FLOG Plants) DOC 02-06-A JFD 3.2-12 ITS SR 3.2.1.1 & 3.2.1.2 Frequency Comment: The ITS SR frequency has been changed from the STS frequency of 12 hours to 24 hours. This is based upon the incorrect justification that the CTS would allow 24 hours based upon ITS SR 3.0.3, since the CTS does not specify a frequency. Adopt the STS SR frequency of 12 hours. FLOG RESPONSE: The change descriptions (DOC 2-06-A & JFD 3.2-12) will be revised to provide a basis for the 24 hours that is predicated on the time required to perform the surveillance. DOC 2-06-A is also revised to be DOC 2-06-M because this change is more restrictive than the CTS. Callaway and Wolf Creek are incorporating this change (DOC 02-06-A, JFD 3.2-12) in lieu of maintaining CTS which did not specify any completion time. DOC 02-13-LG (applicable to Callaway only) and JFD 3.2-17 are no longer used. ATTACHED PAGES: Encl. 2 2-5 Encl. 3A 3 Encl. 3B 3 Encl. 5A 3.2-3, 3.2-5 l Encl. SB B 3.2-7, B 3.2-9
- Encl.6A 3, 4 l
Encl 68 2, 3 l
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. Td24~/C 4.2.2.2 Fo(Z) shall be evaluated to determine if %is within its limit by: ) (a. Verifying F*,(Z) and F*,(Z) satisfy the relationships in the COLR]
^
O. 90!T.; D "0Y _"O!O^^r t.*!^ O ;-J : ' . _ f*j^^
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. "-'4';!n;; M ":!!:::'n; : ':^ - .:.'.:;:
r r'MP tr v/ *ni L' O JL**\**Il r .y) s for V > u.o b dE A 4J r r= MP s r tre mi I A r) II EM E. 11 1."g)s for F $ 0.5 rn <1rwm1 L "' J L ' N /J
=-.xxx-a
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r:T _, _ _ _ ..,'. 'L'; ~_T' ' Di - ~ z'. g[_.a.
-}}_ _ :, :_?_ :, - _:~s:[d AS;E-- MU :- . ' :[5
- h. Measuring Fo"k,(Z) and F",()Z) ccording to the following schedule:
- 1. Once after each refueling prior to THERMAL POWER exceeding 75% RTP, and __
w3 {og.eg,.h n c.. w a + w . .- 2.4 n_.-* ++] ht. , p-- , . ....y equiiipiium csr.sruGns after exceeding, by 10% or of RATED THERMAL PO THERMAL b a s.t.3 l POWER at which , was last determined,* es.1-7 i h2. At least once per 31 Effective Full Power Days, h%thereafter) C
- h. Wit F ,(Z) asurements indicating maximum 0 s K(Z) ,
C '. has increased sina the previous determination of Fo"kF ,(Zheither of the foHowing actions shall be taken:
- 1. Fo" shall be increased over that specified in 4 2.2.2.c by an .w.s n de factor specrfied in the COLR{and reverify F",%
[within hmits)or
*During power esetiatior{following shutdown}et S b;:. "n;; ef eech ~;&. W'M THERMAL POWER may be increased untd a(an equilibrium) power level for extended operation has been achieved after which a power distnbution map may be obtained.
WOLF CREEK. UNIT 1 3/42-5 Amendment No. 64,92 Mark-up ofCTS3M.2 5/l5/97 l t _ _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
J CHANGE BRBER !GC DESCRIPTION COLR. Details of the Fa(Z) measurement, including the treatment of uncertainties, are moved to the BASES. The REQUIRED ACTIONS are re written for consistency with NUREG-1431. The specific changes include the more appropriate use of F/(Z) and F/(Z) versus Fa(Z). 02-02 LS 3 The required ACTIONS are re written for consistency with l NUREG 1431 and industry traveler TSTF 95. The specific changes include the relaxation of the completion time requirement to reduce the High Neutron Flux Reactor Trip Setpoints [from 8 hours] to 72 hours. The reduction of the setpoints is a conservative action for protection against the
- consequences of severe transients with unanalyzed power
! distributions. The completion time of 72 hours is sufficient considering the amount of work required to be done to reduce l the setpoints, the small likelihood of a severe transient in l this time period, and the prompt reduction in THERMAL POWER required upon discovery of the out of limit condition. l 02 03 M The required ACTIONS are re written for consistency with NUREG 1431. The specific changes include the addition of a requirement to be in at least MODE 2 within 6 hours should any j of the ACTIONS not be completed within the required time period. This requirement is more restrictive than the previous requirement to enter LCO 3.0.3, which allowed I hour before the 6 hour shutdown requirement became effective. l 02 04 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 02 05 M Consistent with NUREG-1431, F/(Z) and F/(Z) must be verified to be within limits prior to exceeding 75% RTP after each refueling. This requirement is not explicit in the current TS. The TS are made more restrictive by stating this requirement. 02 06 hM Net WCW See(onvecettIn Go5ar416n,Mhle f, ( o e . 16J sea.T s A - 3a. 02 07 A The footnote allowing the power to be increased until the THERMAL POWER for extended operation has been achieved has been incorporated in the note preceding SR 3.2.1.1 and SR 3.2.2.1 in the ITS allowing power to be increased until an equilibrium power level has been achieved. This footnote replaces the specification 4.0.4 exemption in the current TS. Therefore, the change is administrative, and no technical changes would result. l l WCGS-Description ofChanges to CTS 3M.2 3 5/15M7
INSERT 3A-3a 0 3.2-3/3.2-7 In 'the ITS SR 3.2.1.1 and SR 3.2.1.2, a time limit for assessing Fo(Z) af ter reaching equilibrium conditions is specified. Because the CTS does not have such a time restriction, this change is more restrictive. The time limit for completion of this surveillance, 24 hours following the establishment of equilibrium conditions, has beer selected based on plant experience. Twenty-four hours is a reasonable time for obtaining and evaluating a flux map and then completing the required procedural steps associated with this surveillance. Further, the 24 hour time limit does not allow for plant operation in an uncertain condition for a protracted time period. The time limit of 24 hours is consistent with Amendment No.116 for Wolf Creek in which the NRC approved allowing the performance of a flux map 24 hours after achieving equilibrium conditions from a Thermal Power reduction required with QPTR determined to i exceed 1.02. l l i i I l l l
o L R n - 3 L s s s s i s a fgQ A e e e e oo t e t 1L j e g C Y Y Y Y N n Y a P s b l t n A K E e m e Mo iY E r . iS R uT . Y C qC e T FL R n h't i - I - L O s s s s t s I e e e e oo e Ne B W Y Y Y Y N n Y p A
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Fe(Z) (Fa Methodology) 3.2.1B SURVEILLANCE REQUIREMENTS
.....................................N0TE --- ------ -- --- --- --
During power escalation at the Leginning cf cech cycic following shutdown, THERMAL POWER may be increased until an equilibrium power level has been achieved, at #'3:24U$-- - which a power distribution map is obtained. ' SURVEILUWCE FREQUENCY i SR 3.2.1.1 Verify Q(Z) is within limit. Once after each il refueling prior to THERMAL POWER exceeding 1 75% RTP r M jusn.uo %24 I 431-3 cis 2. -7 _. Oncclithin p D23 EG after achieving ($7M; ' equilibrium 3 .2.-12. conditions after exceeding, by 2 10% RTP, the THERMAL POWER at which Q(Z) was last verified i O 31 EFPD thereafter (continued) 1 WCGS-Mark-up ofNUREG-1431 -ITS 3.2 3,2-3 S/1587
Fo(Z) (Fa Methodology) j 3.2.1B ) SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY qs.2. 3 1 {i% 24 h ]uo $51-7_ t SR 3.2.1.2 (continued) Onc?Uithir. N [12] har:; after achieving
$MP equilibrium 2-a conditions after exceeding, by
- t 10% RTP, the THEPJ%L POWER at which F#(Z) was last verified ANil I 31 EFPD thereafter l l WCGS-Mark-up ofNUREG-1431-ITS 3.2 3.2 5 S/15M7 l
Fa(Z) (Fa Methodology) B 3.2.1 l BASES
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W it4 rib ccnAihnu.ses=Wehda.j* l SURVEILLANCE of Q(Z) and Q(Z) folicwing a _ power increase of more than 102, ensures l REQUIREMENTS that they are verified %sacrop4JpactWa5NMRTP (or any other (continued) level for extended operation)na seneeysso ElMJy_orium condi31onslare atisfeiedMeiEtheMK4ufTfcf_entKetalECR])ll_eM . of_er,@ltgEgohditjens~tofperfereM1sug@g' In the absence of these i Frequency conditions, it is possible to increase power to RTP and I operate for 31 days without verification of Q(Z) and Q(Z). The Frequency condition is not intended to require ~ verification of these i parameters after every 10% increase in power level above the last l verification. It only requires verification after a power level is I achieved for extended operation that is 10% higher than that power at which Fo was last measured. SR 3.2.1.1 ' l l l Verification that $Z) is within its specified limits involves increasing Q(Z) to allow for manufacturing tolerance and measurement l uncertainties in order to obtain $Z). Specifically. Q(Z) is the measured value of Fn(Z) obtained from incore flux map results and
$Z) = Fo$41.6815 ef. 4). $ Z) is %en compared to its specified 11mlu.
M .3. 2. cc - I l l The limit with which $Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR. Performing this Surveillance in H0DE 1 prior to exceeding 75% RTP ensures that the $Z) limit is met when RTP is achieved. because peaking factors generally decrease as power level is increased. If THERMAL POWER has been increased by 210% RTP since the last ! determination of $Z), another evaluation of this factor is required I
- g". / CS12]*;;q after achieving equ111brita conditionsh c1 m q P's c r %w .A t thu hi # a -p r k ;;l (to ensure that Po-(Z) values are xiing reduced sufficiently with power increase to stay within the LCO limits). l_ awks. gy ,)
The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS). (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.2 B 3.2 7 S/1SM7
Fe(Z) (Fa Methodology) B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued) REQUIREMENTS If the two most recent Fo(Z) evaluations show an increase in the expression maximum over z rl(z) I it is required to meet the Fa(Z) limit with the last Pd(Z) increased by a e-t_he apiirTagiff~ tie factor of [1.02] sgiggjed~ifthe COLR, or to evaluate Fa(Z) more frequently, each 7 EFPD. These alternative requirements prevent Fa(Z) from exceeding its limit for any significant period of time without detection. Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the Fa(Z) limit WJ11he met when RTP is achieved, because peaking factors are generally decreased as power level is i increased. Fo(Z) is verified at Dower levels :t 10% RTP above the THERMAL POWER of
- its last verification,'O[12] his after achieving equilibrium cp 3.2. 5 ' conditions Walecuerm.,ymyww6wetp to ensure that i
ca.2.-7 Fa(Z) is within its limit at higher power ' evels. The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of Fo(Z) , evaluations. ! The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, when the ! plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day surveillance. REFERENCES 1. 10 CFR 50.46, 1974.
- 2. R;;ulatory Ouid; 1.77. Rev. O. ".ey 1074 USARESection 1574~.8j i-(continued)
WCGS-Mark-up ofNUREG-1431 - Bases 3.2 B 3.2 9 S/158 7
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CHANGE NUMBER JUSTIFICATION i 3.2 12 Not a c e ( os oit!5rdtT WCff(M-3cL Seed $nverdon Ca$arEondable) g 3.2 13
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This change retains the CTS for the performance or peaking factor determinations following plant shutdowns. The CTS through the exemption to specifit.ation 4.0.4, allows prerequisite plant ! conditions to be obtained prior to requiring that the ! surveillance be completed. {samfe,A-%RQ.s.2 4 3.2 14 This change retains the Wolf Creek CTS for the completion time i for Required Action 3.2.2 A.2. This Completion Time was approved l in License Amendment 61. This change is based on the time required to reduce power, establish equilibritan conditions, and obtain a flux map. 3.2 15 This change incorporates industry traveler TSTF-109. Action A.2 would require the QP1R be determined rather than performing a specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu _of SR 3.2.4.1.)
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44.5.z-1o [ acceptable because they clarify the ISTS regarding frequency and ; use of incore flux monitoring for QPTR measurement. The changes reflect that incore detectors provide an acceptable QPTR determination during all plant conditions. 3.2 16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that is within its limit. Fa(Z) is approximated by i F*a(Z) (which is obtained via SR 3.2.1.1) and F"a(Z) (which is obtained via SR 3.2.1.2). Thus, both F*a(Z) and F*a(Z) must be established to verify Fo(Z). This change is consistent with , traveler WOG 105. 1 ! 3.2 17 EFr ncy r 1remen c perfo n$F q,measur ts b revi to c ora to which not specify a let nT . C ent pr tice i o perfor measurement as s as , racti . The S SR Camp ion Times are sed o what sa no ly rea le Camp ion Time for formi a fl map: ver, i problems cur, the plant be f ed t reduc I power o shutdown. is would subj the nt to tran ent and on witho sufficie/dsafe basis There re, si taining current TS requ ement s accep ble ause i Q useA. ./ V ers.2 .3 7
- WCGS-Iqferencesfrom NUJtEG-1431 - I153.2 3 S/15M7
INSERT 6A-3a 0 3.2-3/3.2-7 The required time for completion of a flux map for determination of the heat flux hot channel factor is changed from 12 hours to 24 hours after achieving equilibrium conditions. The proposed change affects SR 3.2.1.1 and SR 3.2.1.2. Based on plant experience, the proposed time (24 hours) is a reasonable time period for obtaining and evaluating a flux map and then completing the procedural steps associated with this surveillance. Further, the 24 hours time period does not allow for plant operation in an uncertain condition for a protracted time period. INSERT 6A-3b 0 3.2-4 The note was incorporated to address the rare situation where, during a mid-cycle shutdown, through further review of the previous surveillance, it was determined that the surveillance was invalid; or the required surveillance frequency is not met due to the shutdown.. The amended Note would be required to return the reactor to a power level at which a new surveillance could be performed. l i l i \ \ I i i 1
CHANGE NUPEER JUSTIFICATION as apd th AMM ba2M 4.14. ! q ,s.2 3 m- ~_ o 2.z -( woul on e K 3.2 18 This change modifies the requirements in NUREG 1431. Rev.1 for_ Wolf _ Creek to _ retain current TS requirements, i ~ e pavigonJ4o pequgea Ac)sopsyrgposferor y e
.fThe current T5 provisions that would be retained a're (1) the requirement to reduce the Power Range Neutron Flux High trip 1 setpoint during a required power reduction, and (2) the l
restriction that incore flux mapping can be used to verify QPTR i only after one (not "one or more") excore neutron flux channel is inoperable with power greater than 75%. Retention of the current reactor trip setpoint reduction would , add a new Required Action to the QPTR specification. However, the Completion Time would be revised from 6 hours to 72 hours based on industry Traveler TSTF 95. These changes are acceptable because resetting the trip setpoints for QPTR not within limit is based on an assumption that peaking factors are not within limit. The Required Actions of NUREG-1431 also would result in resetting the trip setpoints but only after F and o F% measurements have been taken and detemined to be not within limits. Therefore, : the proposed change is acceptable because it would result in 1 resetting the trip setpoints at an earlier time in the l performance of the Required Actions. The extension of the time period to reduce the flux trip setpoints (frca 6 to 72 hours) is also acceptable because it would permit time to perfom required i l QPTR determination, permit orderly resetting of the high flux ! trip setpoints, and reduce the chances of an inadvertent reactor trip during the required power reduction. The extension of time to 72 hours is consistent with industry traveler TSTF 95. Retention of the incore flux mapping restriction would delete the words "or more" from the Note to SR 3.2.4.2. This change is acceptable because it would prevent operation with less than three excore power range channels available for QPTR measurement l and comparison to the incore flux mapping results. This would provide additional, conservative margin in using the incore detectors to verify QPTR and additional capability for monitoring QPTR between incore maps. 7-lo s.z.c.\
~
The c s bas o acceptabl ause the allow a onal ti r the t conditi to stabil be e rifying ing fa rs. .The ired Acti of G 1431 would provi fficient for the nt t tabil i y . large r reductio to obtai n accu e measur tof) WCGS-Differencesfrom NUREG-1431 - ITS 3.2 4 $/IS/97 I
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-4 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor CTS 3/4.2.3 Nuclear Enthalpy Rise Hot Channel (All FLOG Plants) DOC 02-07-A JFD 3.2-13 SR 3.2.2.1 NOTE and related Bases. Comment: Justify the need for the note related to permitting power ascension after shutdown to a level at which a power distribution map is obtained. It appears that this note is unnecessary, considering the phraseology of the SR Frequency ("Once after each refueling prior Thermal Power exceeding 75% RTP"). Explain the need for this note. The SR 3.2.2.1 Bases also mentions "(leaving Mode 1)" which appears to be the l incorrect mode. I FLOG RESPONSE: The Note, as described in JFD 3.2-13, was incorporated to address the rare situation where, during a mid-cycle shutdown, through further review of the previous surveillance, it was determined that the surveillance was invalid; or the required surveillance frequency is not met due to the shutdown. The amended Note would be required to return the reactor to a power level at which a new surveillance could be performed. The " leaving MODE 1" clarification is based on the Applicability of the LCO (MODE 1, only) and is intended to avoid confusion in a scenario where the plant may be taken off-line (typically, MODE 2), but not
" shutdown" (commonly considered to be MODE 3 or lower).
ATTACHED PAGES: Encl. 3A 3 l 1 l l L_-_________-__________________
~ - -
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CHANGE Nl#BER - --
' JUSTIFICATION 3.2 12 Not c WCfC Seegonvetsfon Casi arMonhe)
( os iW 5ratT M -s a , 3.2 13 as.n -b This change retains the CTS for the performance or peaking factor determinations following plant shutdowns. The CTS, through the exemption to specification 4.0.4, allows prerequisite plant conditions to be obtained prior to requiring that the surveillance be completed. {ses.r c,A-ARQ.3.2.-4 3.2 14 This change retains the Wolf Creek CTS for the completion time 1 for Required Action 3.2.2 A.2. This Completion Time was approved in License Amendment 61. This change is based on the time l required to reduce power, establish equilibritas conditions, and obtain a flux map. . 3.2 15 This change incorporates industry traveler TSTF 109. Action A.2 { would require the QPTR be determined rather than performing a specific surveillance because more than one surveillance can be
~
used to determine QP1R. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2._4.1.)
.i t_f: Y:r ", 2.2.' ? E dr-M t: x;.;re
- 4. f.. -.a if .D
.z___ .._. ..m .;f These cunges are 4 4.5.z-10 [
acceptable because they clarify the ISTS regarding frequency and use of incore flux monitoring for QPTR measurement. The changes reflect that incore detectors provide an acceptable OPTR determination during all plant conditions. 3.2-16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that is within its limit. Fa(Z) is approximated by F*a(Z) (which is obtained via SR 3.2.1.1) and Pa(Z) (which is obtained via SR 3.2.1.2). Thus, both F*a(Z) and Pa(Z) must be established to verify Fa (Z). This change is consistent with traveler WOG 105. 3.2 17 EFr yr ramen or perfo ng Fq,me/ asur ts h , revi to e onn to which not specify a let iT . l C ent pr tice i o perfo measurement as s as racti . The SR Coup ion Times are sed o what sa no ly rea able Como ion Time for formi a f1 map: ver, i problems cur, the plant be f ed t reduc is would'subj power o shutdown. the nt to tran ent cond ion wit suffici safe basis There re, ma aining current requ s accep ble ause i
@ use4. / k 43.2-37 WCGS-Differencesfrom NUREG 1431 -ITS 3.2 3 5/158 7
INSERT 6A-3a 0 3.2-3/3.2-7 The required time for completion of a flux map fo' determination of the heat flux hot channel factor is changed from 12 hours to 24 hours after achieving equilibrium conditions. The proposed change affects SR 3.2.1.1 and SR 3.2.1.2. Based on plant experience, the proposed time (24 hours) is a reasonable time period for obtaining and evaluating a flux map and then completing the procedural steps associated with this surveillance. Further, the 24 hours time period rioes not allow for plant operation in an uncertain condition for a i protracted time period. INSERT 6A-3b 0 3.2-4 The note was incorporated to address the rare situation where, during a mid-cycle shutdown, through further review of the previous surveillance, it was determined that the surveillance was invalid; or the required surveillance frequency is not met due to the shutdown.. The amended Note would be required to return the reactor to a power level at which a new surveillance could be performed. l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-6 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.4 Quadrant Power Tilt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (All FLOG Plants) DOC 04-01-A JFD 3.2-05 ITS Required Action A.5 Comment: The ITS proposes to change the STS wording for Required Action A.5 from
" Calibrate excore detectors to show zero QPTR," to " Normalize excore detectors to eliminate tilt," based upon WOG-95 (and rejected TSTF-25). A preferred wording would be that proposed in the Comanche Peak CTS mark-up, " Calibrate excore detectors to show zero Quadrant Power Tilt." What is status of WOG-957 FLOG RESPONSE: Traveler WOG-95 was transmitted to the NRC in February 1998 as TSTF-241. The FLOG is incorporating TSTF-241 including the latest revisions discussed at the June 1998 WOG MERITS Mini-Group meeting. These revisions corrected errors made during the development of TSTF-241.
Additionally, Wolf Creek submitted a License Amendment Request to CTS 3/4.2.4, Quadrant Power Tilt Ratio, on February 4,1998 which was approved on April 27,1998 in Amendment No.116. This amendment incorporated the changes proposed in TSTF-241. The FLOG believes that it is appropriate to incorporate the proposed TSTF-241 changes based on the NRC approval of the Wolf Creek amendment request. ATTACHED PAGES: l Encl. 2 2-11, 2-11 a, 2-12, 2-13, ; Encl. 3A 5, 6, 7,8 l Encl. 3B 6, 7 , Encl. 4 1,27-35 i Encl.5A Traveler Status page, 3.2-10,3.2-11,3.2-12 Encl. SB B 3.2-25, B 3.2-26, B3.2-27 Encl.6A 1, 4, 5 ! Encl. 6B 1, 3 l O
POWER DISTRIBUTION LIMITS -
~ -
EgyE.pu. CT S 3/4. 2. 3/4 2.4 OUADRANT POWER TILT RATIO ufsh INSERT f.-Il Q3.Z-le ) LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRAN POWER TILT RATIO shall not exceed 1.02 APPLICABILITY: MODE , above 50% of RATED THERMAL POWE ACTION:
- a. With the QUADRANT WER TILT RATIO detennined to exceed 1.02-but 13~
' -- ?- : :_ 'to t.. 2 > -.:m e C 9 ' *-- O U.^S P. PO'f." v!LT o^7?O et '-- ' .00 p-- hour r "' 2:n . . .
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04i1, 0-LS ~14? J m.
- ) ,: "__et.,- 5: OU^SP.^.Sm oO'f._ v!LT P.^7!O te .J'"-
@b) Reduce THERMAL POWER atlea 3% from RATED THERMAL R PO for each 1% ofindicated OUADRAN POWER TILT RATIO in exce of 1 04-10-I$14 ~ '
and[within 72 hours after each QPTtt determinationlsimilarly r ce the ~ ~ ' - - ' Power Range Neutron Flux-High Trip pn' z^^2 . !M nedri F =~' * } b) At least once per 12 hours, determine QUADRANT R TILT RATIO, and c) Within 24 hours after achieving equilibri conditio a THERMAL 0010-L&l4 ~ POWER reduction required by Action a.1.a), nd onc per 7 days thereafter, A '2-confirm that the Heat Flux Hot Channel Fac Fo( is within its limit by performing Surveillance Requirement 4.2.2.2 d onfirm that Nuclear Enthalpy Rise Hot Channel Factor, Fh,is withi limit by performing Surveillance Requirement 4.2.3.1.;
- 2. Prior to increasing THERMAL POWER and P Neutron
- .061MC Flux High trip setpoints above the limit of on a.1. -
.ansil a) Re. evaluate the safety analyses and onfirm that uits remain valid for the duration of operation r this conditi , and only then b) Normalize excore detectors to el' inste tilt; 5Yk 3.* Within 24 hours after achieving uilibrium conditions after ching RTP or within 48 hours after increas THERMAL POWER above limits of - "04-INLflU
~"
ACTIONS a.1.a), confirm that Z)is within its limit by performi Surveillance Requirement 4. 2.2 and that Fh is within its limit performing Surveillance Requirement .2.3.1; and RA E MAL ER in no 4 hou
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"W~ ACTION a.3 must bhompleted when ACTION a.2.b)is implemented.)
I WOLF CREEK - UNIT 1 3/4 2-11 Mark-up of CTS 3H.2 S/lS/97
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION AC. TION ntinued) iG 3,g,4 1 '.!:9 *e O a
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POVER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Contin ') Q 3 7.-la k.
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Ot.%1 A ED.. .. I 1 l l l WOLF CREEK UNIT 1 3/4 2-12 Mark-up ofCTS3N.2 S/lS/97
- _ _ _ _ _ _ _ _ _ - _ _ _ _ _ = _ _ _ _ - _
I l POWER DISTRIBUTION t IMITS ! LIMITING CONDITION FOR OPERATION ACTION (Continued) I'N l l o o.a.. Tuema An enuco i....u enu ms o ATen vuentui --' onure
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... w : :. ..$. . A,4 SURVEILLANCE REQUIREMENTS 4 2.4.1 The QUADRANT POWER TILT RATIO s 11 be det ined to be within the limit above 50% of RATED THERMAL POWER b
- a. Calculating the ratio at least once per 7 days . . _- t: !:- :: 6fMlF@
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r----- ----'-----'-'-r-'-- - 4.2.4.2 The QUADRANT POWER TILT RA O shall be det ined to be within the limit when above 75% of RATED THER POWER with o Power Range Channel inoperable by using the movable incore tect
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90WiiR-T4kT-RAT 40 at least once r 12 hours. {M**M64~-blICdlil j i 1 l l i WOLF CREEK - UNIT 1 3/4 2-13 Mark-up ofCTS3M.2 5/15/97
l I POWER DISTRIBUTION LIMITS Ud 3/4 2.4 00ADRANT POWER TILT RATIO j LIMITING CONDITION FOR OPERATION 3.2.4 The OVADRANT POWER TILT RATIO shall not exceed 1.02. l'*l-A APPLICABILITY: MODE 1. above 50% of RATED THERMAL POWE ACTION: l
- a. With the OUADRANT POWER TILT RATIO determined to exceed 1.02 l !
- 1. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for i :
each 1% of indicated 00ADRANT POWER TILT RATIO in excess of l 1.00 within 2 hours after each OUADRANT POWER TILT RATIO determination and
- 2. Determine the OUADRANT POWER TILT RATIO at least once per 12 hours, and
- 3. Within 24 hours after achieving equilibrium conditions from a THERMAL POWER reduction required by ACTION a.1. and once per 7 days thereafter:
a) Confirm that the Heat Flux Hot Channel Factor F (Z). is within its limit by performing Surveillance Req,uirement , 4.2.2.2 and j b) Confirm that Nuclear Enthalpy Rise Hot Channel Factor. F", is within its limit by performing Surveillance Requirement 4.2.3.1. and
- 4. Reduce the Power Range Neutron Flux-High Trip Setpoints a 3%
for each 1% of OUADRANT POWER TILT RATIO > 1.00 within 72 hours after each OUADRANT POWER TILT RATIO determination, and
- 5. Prior to increasing THERMAL POWER and Power Range Neutron Flux-High Trip Setpoints above the limits of ACTION a.1. and a.4. .
reevaluate the safety analyses and confirm that the results remain valid for the duration of operation under this condition, and
- 6. Prior to increasing THERMAL POWER above the limit of ACTION a.1.. normalize excore detectors to restore OUADRANT POWER TILT RATIO to within limit and l
l l
*Sr Spri:' T=t S;;ptica Sprificticr. 3.10.2. :-oi-A l
WOLF CREEK - UNIT 1 3/4 2-11 Amendment No.116
149JtT 2-11 Q 3.7 -(. { POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) l
- 7. Within 24 hours after echieving equilibrium conditions not to exceed 48 hours after increasing THERMAL POWER above the limit of ACTION a.1.**:
a) Confirm that the Heat Flux Hot Channel Factor F (Z). is within its limit by performing Surveillance Req,uirement 4.2.2.2 and b) Confirm that Nuclear Enthalpy Rise Hot Channel Factor. F % is within its limit by performing Surveillance Requirement 4.2.3.1. I
- b. If the requirements of a.1.. a.2.. a.3.. a.4.. a.5. a.6 . or a.7.
above are not met, reduce THERMAL POWER to s 50% of RTP within the-next 4 hours.
- c. W e ~ evisier.: cf Spaci+1 =ticr. 3.0.4 are r.;t~;ppli a bic. 4-61-A l SURVEILLANCE REQUIREMENTS i
w.3.z.co i l - 4.2.4.1 The 00ADRANT POWER TILT RATIO r.haL1 be determined to be within t limit above 50% of RATED THERMAL POWER by:lfD 4-of-A l 4-II-A l
- a. Calculating the ratio at least once per 7 days 2= th: :1;r: is oi-o1 to.
OPEPX LE. onu 5 5. h 5 t N. ; b -*We 4.2.4.2 TheOUADRANTPOWERTILTRATIOshallbedeterminedtobewithinihe' limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors t: : = fir; th;t th; 443 L(c nordird ;;;tric p;w;r distributicr.. Obt:ir. d fr;;istwo ;;ts cf f;;r E x.;tric thimble lui.etiona vi e Tull vvi c ilua ..op, censistent ,,ah the "d4ceted OL'TP "T POL'EPs TILT RATIC at least once per 12 hours (++ 4-in- A] w. uoll R-tstk in Mrom .n %.r Rage. Neuba Fla. channel inopemb l f an4.TdanAt. potaes. t. 7s% are, h remainin3 +hre.e. paar CNil rwy- 4 LohanneAs can be. u -A for c.tcaista4 onn.. j 4-u-Aj {sw. 4.2 4.s. my be. perkned A ha of h suv vstWu-. t _** ACTION a.7. must be completed when A_CTION a.6. is performed.
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- 3
- ll
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emosa. n5% arf. , WOLF CREEK - UNIT 1 3/4 2-12 Amendment No.116 (Next page is 3/4 2-14)
CHANGE 1 NUMBER EliC DESCRIPTION 03 02 LS-5 Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 38). ; 1 03 03 M The Requirement to reduce power to less than or equal to 5% RTP (exit Mode 1) within the next 6 hours is added in lieu of the use of LCO 3.0.3. This requirement is more restrictive than the previous requirement to enter LC0 3.0.3, because LCO ; 3.0.3 allowed 1 hour before the 6 hour shutdown requirement became effective. 03 04 LS 6 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 03 05 H If the enthalpy rise hot channel factor action statements requiring flux mapping and correction of the cause or power reductions are entered, they must be completed, even if compliance with the LCO is restored. These requirements from NUREG 1431 are more restrictive than the corresponding requirements from current TS. 03 06 A Consistent with NUREG 1431, a note would be added to state that THERMAL POWER does not need to be reduced below the power required by Action a. in order to comply with the series of flux maps required by Action c. This is a clarification of the current TS in that if compliance with the LCO is restored prior to reducing power level below 50%, flux maps need only be performed for those plateaus traversed. If power level did not drop below 95% no flux map would be required. No flux map would be required by Action c, but would be required by Action [b]. 03 07 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 03 08 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 03 09 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 1 03 10 LG Not applicable to WCGS. See Conversion Comparison Table
- (Enclosure 3B).
L e _ - 4 3.Z- G. I 04 01 A [Clarifie that w the exc detector are cal rat , th . l Quadr P ilt is oed out. he QP s no liz to u y.) is requir nt from EG 1431 mod ed b ST - 1
- 5. i consister with the rent TS ION r uir ts or grer fying QP is within mit durin power scala ~ n WCGS-Description of Changes to CTS 3M.2 5 5/15/97
CHANGE MREB NSE DESCRIPTl0N
-es.I-Q s ' id!nt>f91nr1fnd sofrectThgAt1is cat (s'e.erQPTR.aut f m . & otylicalAA.h M S.5** G n d c.' mpar 4eTabj s (Enclosww 35),
04 02 LS 10 ' red tion cal ula once hour [The QP il L wa edu t ess n 50% wo be imi a rep ed new uir s fr NUREG 431. Thi epr ts r ion requir nts f moni ri nd r uci . propo chan would educe fr ene f calcu ion: ver, is is cc ble au cha s in QP foll ng the equir r ction i wi r atively ow a he nor freq ca ula oy" requir by cur TS ld div the .t - no cofftrol staff f corr ve a n acti tie(.144 ar wcces.. see conver4m comew%n Tateta. ar nc.t.ghcabt.a. mss) 04 03 LG The details regarding obtaining QPTR using the incore detectors would be moved to the BASES. This is an example of removing details that are not appropriate for inclusion in TS. Thus, the proposed change is acceptable. 04 04 LS 12 Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 38). 04 05 LS-11 Not applicable to WCGS. See Conversion Comparison Table 43.2-6,] (Enclosure 38). St. 9ncaw. 4=, wccrs.se,c o,nvwden C capmirnTaa.N* "M. 4 06 LS 13 Actions volvi TRs 1.0 woul el nated ,in N conf i nce 1431 Whil he r ire r rdi TRs i exces of 1. due misavi' p ~.-sof yS cont rods d be ddres by ITS uir ts a ciated hr group sali nt its, CT ctions regardi TRs n exce of 9 due o ot caus would be i 4 repla by ss res ctiv requi ts The requ e t the be cula onc r r and t be i o less n RTP thin hours nd the ower lux gh tr set nt be Ran Neut uced thin t l l ng 4 . I addit' n, t Sr ire i tifica onand{ i_ l corr:ectn of ca e of tilt nditi and odic 3 veri cati that is ithin '1mits ing a subsequent j I (acalcunsioedtooRy oncThe r1 . S d requ rs, (1) t t QPTR
) only 3% RTP uction f
t for ach of QP in ess of .0 and redu n in fl f to a foll ng ps ints and verif ation o peaki actors p or r asc sion a a reeva ation of afety a yses ior power scensio Howev , the r r nts
) ft Sa accep le bec e (1) t QPTR wo d (
e ed cha lowly o r time a less f I cu ion of woul acce ble: (2) nce o ting s f come es a reducti ,i cco[rnce 1th ITS M uiremen , the ef ect of an fl tiltMfill tend WCGS-Description of Changes to CTS 3N.2 6 5/15/97
CHANGE NutBER H2iG DESCRIPTION . fos.2.-Gl l . --_ to be i ed y educ g he f d es blis _ nggpateM ma t f ign a s, t red tion f r ui t IT r it na nt t nsie that ra uld s ere han red tion o le th t - uir CT fur r, e mina t tri et nt r on ept e ause OPTR n exc so imi k ssa y1 yt acci nt a yse ass ti I av vio ed: (3) I Requi ion ri o
.s seg t to r ensio prov e as ca th pper
)o atio at or ar R will in cord ce wi h t sdfetyj u
ly s and, heref e, ac tabl g
~
i 04 07 A The statement that Specification 3.0.4 does not apply is no 4 longer needed as revised Actions permit continued operation for an unlimited period of time. In accordance with the provisions of ITS LCO 3.0.4. this is acceptable. 04 08 Not used. 04-09 A St ;;'tidi to G . 1 Cw..m u w C w arib Mi WC S 1-e01) ( (En rr; 'Q INSE8T 3A-7j h used .T -
@M 04-10 The r at cha s ) r ise the -2M 8 y e on r ett t Ra on F1 \
I ri set n dur t re tion ollow QPTR su , wo d c1 ify t rr i must
>c et w in r aft ea tenni tion I a evi eq ibri co ition for suri aking !
f tor a (3) ld imi te t curr TS S l r ri QP o est ed wi in 2 urs, to ri d ng etu to r. rese ng the r R ron ux igh ip s int < 55% lowi powe edu on 50% or low. urrectE requi a rr cti wit 2 s of terni TR of limit l and igh ux p se int uctio ithin next l i rs. leti ti or res ing t high f trip set ts uld chan to 7 urs er dete ning TR l out lim Th is ac tabl cause would orderly f
. rait i ti o pe rm r red de inat1on]persi 3, '
i setti of t igh ux tr setpojits,and uce y cha s of 'inad tent actor trip duri the r ired rr ctio Eli ating t/curren S ACT S requ ing
\ be tor ithin3+' hours, to verifi
/ ing eturt o power,/and res ing t ower R I I Neu n Flu - High rip set to < foll g r uctio o 50% or is ac table; aus , they are cons ent wi G- 1. Rev. and a ete in N i LS 4, the JT$ req ements ld pro e acce able S Actions that are pro sed for WCGS-Description of Changes to CTS 3M.2 7 S/15/97
CHANGE NUISER EilE DESCRIPTION m.g l
'n. T roposed hanges,bai n traveler 95, flimina whi clarif at reductions be compl withi 2 rs a each determina and persi chievi
, equ rium c itions for suring peak factor re npt ' l nsider o be relax ns of curr equir s.gs is l beca completion s for thes ctivitie re not
=c -
_ specified in cu ent TS./ l fo4 -il A. 94 san' sA eI 1 we. s. a. -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). gcP 3.1.-ool T
~
! 05 03 LG Co sten ith EG 14 , the equ fo a l E I ION he RC se rs t1 st nce
' 18 hs the irene t no 11 t c 1 ae I ed the s for t RCJ/ fl -
r cto tr f on in Secti 3.3 A. riot opplic2lAtintaus5. l Esse cever.fien chmperr.MTaMa Catnc.n== 3s 05 04 LG Consistent with industry traveler TSTF 105, the explicit requirements that the RCS flow be measured through the use of a precision heat balance measurement and that the instrtamentation used in the performance of the calorimetric flow measurement be calibrated within a specified time period of performing the measurement is moved to a licensee controlled document. The requirement to verify that the RCS , flow is within limits remains within the Technical ; Specification. This is an example of removing unnecessary i l> details from the TS and is acceptable based on the guidance provided in 10 CFR 50.36.
~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.
l' 05 06 LS E 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 DNS related l parameters of Reactor Coolant System (RCS) average temperature, pressurizer pressure, and RCS flow rate are WCGS-Description of Changes to CTS 3N.2 8 5/15/97
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i NO SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) CONTENTS I. Orga ni za t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 II. Description of NSHC Eval uations. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 ; i l III. Ger'"ic No Significant Hazards Considerations I "A" A&iini strati ve Changes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 "R" - Relocated Technical Speci fications. . . . . . . . .'. . . . . . . . . . 7 "LG" - Less Restrictive (Noving Information Out of the Technical Speci fications) . . . . . . . . . . . . . . . . . . . . . . . . 10 "N" - Nore Restrictive Requirements....................... 12 IV. Specific No Significant Hazards Considerations "LS" LS 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS 2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS 3...................................................... 15 LS 4...................................................... 18 LS 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS 8...................................................... 21
- LS 9.................................. . . .. .4
! LS 10................................. *
.___W."c._".*._._........ -Io u-Il LS 11. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e LS-12.................................. _
.... licable LS - 13. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . M.*reO fe.4Afr . . . . . .
LS 14. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . St # M4r . . . . ict s.6 -2. ) LS 15. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e i l WCGS- NSHCs - CTS 3N.2 1 S/158 7
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IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS l I NSHC LS 10 10 CFR 50.92 EVALUATION l FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS i l The required current Technical Specification Action to calculate QP once per hour until THERMAL POWER was reduced to less than 50% RTP would be eli nated and replaced by new requirements from NUREG 1431. This represents reduction in requirements for monitoring and reducing power. With QPTR no within limit, the CTS require a calculation of QPTR at least once per hour until ther the QPTR is restored or THERMAL POWER is reduced to < 50% RTP. The arable ITS actions would require OPTR to be calculated at least once per 12 houre and continue to reduce THERMAL POWER by at least 3% for each it that QPTR exc s the limit until either QPTR is restored to within its limit or 50% RTP is a ieved. The 12 hour frequency is sufficient because, as stated in the NUREG 1431 ases, further changes in QPTR would be relatively slow. The once per hour fr ency is excessive considering the slow rate of flux change and potentially would ivert the attention of control room staff from corrective action with respect to TR. This proposed TS change has been evaluat and it has been determined that it j involves no significant hazards consider ion. This determination has been l performed in accordance with the criter a set forth in 10 CFR 50.92(c) as quoted below: 1 "The Comission may make a inal determination, pursuant to the procedures in 50.91, that a proposed a ndment to an operating license for a facility licensed under 50.21 (b or 50.22 or for a testing facility involves no significant hazards c sideration.1f operation of the facility in accordance with the proposed a dmert wculd not:
- 1. Involve a gnificant increase in the probability or consequences of an accident reviously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any acci nt previously evaluated; or
- 3. I 01ve a significant reduction in a margin of safety."
The followi evaluation is provided for the three categories of the significant hazards c sideration standards:
- 1. s the change involve a significant increase in the probability or consequences of an accident previously evaluated?
proposed change involves only the compensatory measures to be taken should the TR be outside its limit. The frequency with which QPTR is calculated is not assumed in the initiating events for any accident previously evaluated. In WCGS- NSHCs - CTS 3M.2 27 S/258 7
%31-Gj IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 10 (continued) addition, the change does not involve any new operating activities or hardw changes. Therefore, the proposed change would not significantly increase probability of an accident previously evaluated.
Once THERMAL POWER has been reduced appropriately in proportion to t amount that QPTR exceeds 1.0, any additional change would be sufficiently slow at a 12 hour interval for recalculating QPTR will provide an adequate level of rotection. Therefore, the proposed change will not significantly increase consequences of any accident previously evaluated. l
- 2. Does the change create the possibility of a new or ifferent kind of accident '
from any accident previously evaluated? The proposed change does not involve a physical alt ation to the plant: no new or different kind of equipment will be installed. A o, the manner in which the plant would be operated would not be altered. Thus, change will not create the possibility of a new or different kind of acc nt from any previously evaluated.
- 3. Does this change involve a signifi reduction in a margin of safety?
The proposed change will continue to sure that the plant is maintaired in safe condition while QPTR is in excess of ts limit. Additionally, calculating QPTR once per 12 hours as opposed to every r while QPTR is in excess of its limit would avoid the diversion of personnel esources from corrective actions with regard to meeting the LCO. Therefore. t proposed change will not involve a significant reduction in any margin of s ty. Based upon the preceding fonnation, it has been determined that the proposed change to the Technical ification does not involve a significant increase in the probability or con es of an accident previously evaluated, create the ; possibility of a or different kind of accident from any accident previously j evaluated, or in ve a significant reduction in a margin of safety. Therefore, it l' 1s concluded t the proposed change meets the requirements of 10 CFR 50.92 (c) and does not invo e a significant hazards consideration. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION i Bas n the above evaluation, it is concluded that the activities associated with "LS-10" resulting from the conversion to the improved TS format satisfy the no gnificant hazards consideration standards of 10 CFR 50.92(c): and accordingly, a significant hazards consideration finding is justified. WCGS-NSHCs - CTS 3M.2 28 S/158 7
ciclete,-nd gh0N j d@ 2-b } ; l IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS n NSHC LS-13 l 10 CFR 50.92 EVALUATION l FOR ! TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS ! 1 Actions involving QPTRs of 1.09 would be eliminated in conformance w h NUREG 1431. l While the requirements in CTS regarding QPTRs in excess of 1.09 d to misalignment l of control rods would be addressed by the ITS requirements assoc ted with rod group l misalignment limits, the CTS Actions regarding QPTRs in excess f 1.09 due to other causes would be replaced by less restrictive requirements. CTS require that the QPTR be calculated once per hour and that power be reduced o less than 50% RTP within 2 hours and the Power Range Neutron Flux High trip setpoint be reduced within the next 4 hours. In addition, the CTS require identif ation and correction of the l cause of the tilt condition and periodic verification hat QPTR is within limits i j during any subsequent ascension to RTP. The ITS wo d require (1) that QPTR be i calculated only once per 12 hours (2)~only a 3% reduction for each it of QPTR j in excess of 1.0 and no reduction in flux trip s points, and (3) verification of l peaking factors prior to and following power a ension and reevaluation of safety
- j. analyses prior to power ascension.
The purposed changes are acceptable becau : i (1) the QPTR would be expected to e nge slowly over time so a less frequent calculation of QPTR would be ceptable: l (2). once the operating staff c nces a power reduction in accordance with ITS f requirements, the effect f any flux tilt will tend to be mitigated by reducing the flux and tab 11shing greater margin to fuel design limits, 1 . l the reduction of r required by the ITS would result in a plant transient that generally d be less severe than the reduction to less than 50% as j required by CTS and ; elimin'ating trip setpoint reduction is acceptable because a QPTR in excess of limits s not necessarily imply that accident analyses assumptions have l been vio ted; and l l l
-(3) the I Required Actions prior to and subsequent to power ascension provide ;
[ ass ance that power operation at or near RTP will be in accordance with the ! s ety analyses and, therefore, acceptable. l This roposed TS change has been evaluated and it has been determined that it in lves no significant hazards consideration. This determination has been performed accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below: WCGS- NSHCs - CTS 3N.2 29 SMSM7 '
IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-13 (continued)
"The Conr11ssion may make a final determination. pursuant to the ocedures in 50.91 that a proposed amendnent to an operating license for a cility licensed under 50.21 (b) or 50.22 for a tasting facility inv ves no significant-hazards consideration, if operation of the fac 1ty in accordance with the proposed amendnent would not:
- 1. Involve a significant increase in the probbbil y or consequences of an accident previously evaluated; or
- 2. Create the possibility of a new or diffe nt kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in margin of safety. "
The following evaluation is provided for the ree categories of the significant hazards consideration standards:
- 1. Does the change involve a signifi nt increase in the probability or consequences of an accident pre ously evaluated?
The proposed changes do not involv any new methods of operating the plant or hardware changes; thus, the propo change has no effect on the probability of an l accident. The proposed change involve only the compensatory measures to be taken should the QPTR be outside its limit. These compensatory measures are not assumed in the initiating events for an accident previously evaluated. Therefore, the proposed change will not affect probability of any accident previously evaluated. Furthermore, the p ed changes to these compensatory measures, which are derived l from NUREG 1431. d continue to provide acceptable levels of protection. l Therefore, there w 1 be no effect on any of the accident analysis assumptions and the consequences f the accident analyses are unaffected by this change. l l Therefore, t proposed changes will not significantly increased the probability or consequence of any accident previously evaluated. I
- 2. s the change create the possibility of a new or different kind of accident rom any accident previously evaluated?
The proposed changes do not involve a physical alteration to the plant: no new or WCGS- NSHCs - CTS 3M.2 30 S/153 7
de-ebphcMa. -l ce 3.z.Q IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS t i NSHC LS 13 ) (continued) ) different kinds of equipment would be installed. The changes would n alter the manner in which the plant operated. Thus, the changes would not er te the possibility of a new or different kind of accident from any acci t previously evaluated. The proposed change involves only the compensatory measures o be taken should j the QPTR be outside its limit. The assumptions of the a . dent analyses are unaffected by the proposed change. No new permutation or event initiators are l introduced by the proposed alternate methods of deali g with OPTRs in excess of 1.09. Therefore, there is no possibility for a n or different kind of accident.
- 3. Does this change involve a significant r uction in a margin of safety?
The proposed changes, which involve imp 1 ting NUREG 1431 requirements, will continue to ensure that the plant is mai ained in a safe condition wihtin the envelope of the safety analyses while is in excess of its limit. While differenct actions are taken ins res se to a QPTR in excess of 1.09, the proposed changes would assure that accident alyses assumptions continue to be met. Therefore, the proposed changes w 1 not involve a significant reduction in any margin of safety. l Based upon the preceding in rmation, it has been determined that the proposed I change to the Technical S ification does not involve a significant increase in the probability or conseque es of an accident previously evaluated, create the possibility of a new o different kind of accident from any accident previously evaluated, or involv a significant reduction in a margin of safety. Therefore, it is concluded that proposed change meets the requirements of 10 CFR 50.92(c) and does not involve significant hazards consideration. NO SIGNIFICANT HAZARDS CONSIDE ETION DETERMINATION Based on above evaluation, it is concluded that the activities associated with NSHC "L 3' resulting from the conversion to the improved TS format satisfy the no signif' cant hazards consideration standards of 10 CFR 50.92(c); and accordingly, a no s nificant hazards consideration finding is justified. l WCGS- NSHCs - CTS 3N.2 31 S/2SM7
tyk UMA, a3.2-G IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS 14 : 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The proposed QPTR related changes would (1) revise the completion time for resetting the Power Range Neutron Flux High Trip setpoints during the power red on following QPTR measurements (2) clarify that power reductions must. completed within 2 hours after each QPTR determination and permit achieving ilibrium conditions for measuring peaking factors, and (3) eliminate the c rent TS ACTIONS requiring QPTR to be restored within 24 hours QPTR to be verifi during a return to power,- and resetting the Power Range Neutron Flux-High trip etpoint to < 55% following a power reduction to 50% RTP or below. The propos changes which clarify that power reductions must be completed within 2 hours aft each QPTR determination and permit achieving equilibrium conditions for measuri peaking factors are based on traveler WOG 95 and are considered to be clarificat ns rather than relaxations of current requirements. This is because completion imes for these activities are not specified in current TS. Therefore, those cha s are not evaluated in this NSHC determination. Current TS require a power reduction within rs of determing QPTR out of limit and a high flux trip setpoint reduction wit n the next 4 hours. The completion time for resetting the high flux trip set ints would be changed to 72 hours after determining QPTR is out of limit. This hange would be consistent with industry , traveler TSTF 95 which would extent t allowed time to 72 hours for reducing Power l Range Neutron Flux-High trip setpoi s when peaking factors are out of limit. l With QPTR not within limit, cur t TS require the THERMAL POWER and the Power Range Neutron Flux High reactor tri setpoints to be reduced. Two hours are allowed for , power reduction, and an add ional four hours are allowed for the completion of the ! setpoint reduction. A etion Time of 72 hours to reduce the trip setpoints, as , proposed, will ' allow ti to reduce reactor power, perform the required QPTR determination, and pe t an orderly resetting of the nigh flux trip channel setpoints while red ng the chances of an inadvertent reactor trip during these evolutions. Duri the trip setpoint change, there is increased potential for human error resulting a plant transient. In addition, the reactor power would be reduced within hours; this would provide additional margin to fuel design limits. Finally the ability of an event occurring during the 72 hour period prior to the reduction " high flux trip setpoints is small. Therefore, a completion time of 72 hours for a required setpoint change is justified. Curr TS Actions requiring QPTR to be restored within 24 hours or reduce power to
<5 RTP and requiring verification of QPTR during return to full power operation 1d be eliminated in accordance with NUREG 1431. Also, the requirement to reset the power range neutron high flux trip setpoints after a required reduction to s 50t WCGS- NSHCs - C153M.2 32 $/15/97 i
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73.2-Q IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-14 (continued) RTP would be eliminated. The ITS would add new Required Actions for QP out of limit including requirements for measuring Fo(Z) and P, and performing afety analyses to verify peaking factors are acceptable prior to and foll ng a return to power. The ITS would require a re evaluation of the safety analys prior to increasing reactor power above the reduced power required by the TR limit. Finally, the ITS would require a confirmation that Fa (Z) and P re within limit following the power increase. The ITS focus on maintaining the peaking factors Fo (Z) a P, within limits rather than the QPTR. This is appropriate because QPTR is a itored parameter that is indicative of peaking factor problems. The ITS, as ified by traveler WOG 95, require verification that Fa (Z) and P, are within 1 its within 24 hours after achieving equilibrium conditions by performing SR that can directly measure flux shapes in the core. If Fe(Z) or P, are not wit n limits, the Conditions for those TS will specify additional required actions. ince the peaking factors are of prime importance, the ITS will ensure that the distribution remains consistent with the initial conditions assumed in safety a lyses. The ITS would retain the 2-hour requir t to reduce power proportionally to the percent that QPTR exceeds its limit, is would result in a power reduction that would provide additional margin to 1 design limits during a flux tilt condition to assure that design limits are t challenged by local flux peaking. These design mergins are set conservatively a provide further assurance that operation in accordance with the Required A ions during or beyond the 24 hour period would not challenge fuel design limits The proposed changes _also ld eliminate the requirement to reduce the setpoints'to i
.s 55% RTP within 4 hour of reaching 50% RTP would be eliminated. This change is acceptable on the basi that the ITS would still required the Power Range Neutron
- Flux - High trip set ints to be reduced during the power reduction as discussed above: and the acc ablility of changing the Completion Time from four to 72 hours has been address previously.
I L This proposed changes have been evaluated and it has been determined that they involve no s ificant hazards consideration. This determination has been performed l in accorda e with the criteria set forth in 10 CFR 50.92(c) as quoted below: e Connission may make a final determination. pursuant to the procedures in 0.91. that a proposed amendnent to an operating license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration. If operation of the facility in accordance with the proposed amendnent would not: 1 WCGS- NSHCs - CTS 3M.2 33 5/15/97
)
l
l Q 3.2.-6, } w_ _ _ . _J IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-14 i (continued)
- 1. Involve a significant increase in the probability or conse ences of an l accident previously evaluated; or 2 Create the possibility of a new or different kir ' of cident fran any accident previously evaluated; or I 3. Involve a significant reduction in a margin of s ety. "
The following evaluation is provided for the three categori s of the significant hazards consideration standards:
- 1. Does the change involve a significant increase 'n the probability or consequences of an accident previously evalu ?
The proposed change involves only the compensat y measures to be taken should the l QPTR be outside its limit. These compensatory asures are not assumed in the initiating events for any accident previous 1 evaluated. Therefore, the proposed change will not affect the probability of y accident previously evaluated. Furthermore, the proposed changes to the compensatory measures, which are derived from NUREG-1431 (as modified by industr traveler), would continue to provide acceptable levels of protection. The fore, the proposed changes will not significantly increase the conseque es of any accident previously evaluated.
- 2. Does the change create the ssibility of a new or different kind of accident from any accident previo ly evaluated?
The proposed changes do not nvolve a physical alteration to the plant: no new or different kinds of equipme would be installed. The changes would not alter the manner in which the plan would be operated only the timing of actions that provide potential mitigation of accidents. Thus, the changes would not create the possibility of a new different kind of accident from any accident previously evaluated.
- 3. Does this ange involve a significant reduction in a margin of safety?
The proposed anges will continue to ensure that the plant is maintained in a safe condition hin the envelcpe of the safety analyses while CPTR is in excess of its limit. T refore, the proposed changes will not involve a significant reduction in any marg of safety. Bas upon the preceding information, it has been determined that the proposed cha to the Technical Specification does not involve a significant increase in the p bability or consequences of an accident previously evaluated, create the ssibility of a new or different kind of accident from any accident previously WCGS-NSHCs - CTS 3M.2 34 S/1587 i _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ .l
- 1 dAltde.-M tuA G 3. Z-l. '
l IV. SPECIFIC NO SIGN 1FICANT HAZARDS CONSIDERA S NSHC LS 14 (continued) evaluated, or involve a significant red on in a margin of safety. Therefore, it l 1s concluded that the proposed cha ts the requirements of 10 CFR 50.92(c) and I does not involve a significant h ds consideration. NO SIGNI HAZARDS CONSIDERATION DETERMINATION Based on the above aluation, it is concluded that the activities associated with NSHC "LS-14" re ting from the conversion to the improved TS format satisfy the no significant ards consideration standards of 10 CFR 50.92(c); and accordingly, a no signif ant hazards consideration finding is justified. l l l WCGS-NSHCs - CTS 3N.2 35 y,ggy l
r Industry Travelers Applicable to Section 3.2 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF 24 Not Incorporated NA Not NRC appr3ved as of traveler cutoff date. l TSTF-95 Incorporated 3.2-06 Approved by NRC. l l TSTF-97 Incorporated 3.2-07 Approved by NRC.
^
! TSTF-98, Rev.1 Incorporated 3.2-03 l TSTF-99 Incorporated 3.2-08 Approved by NRC. l l TSTF-109 Incorporated 3.2-15 Approved by NRC. TSTF-110, Rev.(h Incorporated 3.2-10 i TE M: 24) TSTF-112, Rev.1 Not Incorporated NA Not NRC approved as of l traveler cutoff date. l TSTF-136 Incorporated NA {p p w % s.2-00 4 ] TSTF-164 Incorporated 3.2-11 Applicable to CAOC only l l (CPSES). rary,gg' ,v -- -
^ --
%' s.t-(.\ l
. . . . , , Incorporated 3.2-05,9:9-HF Willincorporate portions j gr4 Prf.L _
s . 2. - o9 ofTSTF-25.
- WOG-105 Incorporated 3.2-16 l
l l f i l l
QPTR 3.2.4 l 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR) I LC0 3.2.4 The QPTR shall be s 1.02. APPLICABILITY: H00E 1 with THERHAL POWER > 50% RTP. ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME A. QPTR not within limit. A.1 Reduce THERMAL POWER 3.2-o 5 2 hours 2 tte ; 2 3% from RTP for each gg3E(EIR lt of QPTR > 1.00. ettg@tjprj g 43.1-6 A.2 Perfer;;; L", 3.2.4.1 Once per Determittel.!EIB-end 12 hours pg]. n M " ", ." . "" " p.
)?,I., .s . LQ. .,"'?, If.[. =*'
1 M l qn-6 ) ' A.3 Perform SR 3.2.1.1.LSR 24hoursaRer. #iFits165 3.'2.l.2 and SR 3.2.2.1. ac_h1gyLng _ l eauilibrilm gg] cmRmanorX t!EIERHAL piker
-mon i28GeOftf ter%3.2.I,,
IB E lreg p lort M1 j t M \ Once per 7 days { thereafter l 1 1 M (continued) WCGS-Mark-up ofNUREG-1431 -ITS 3.2 3.2-10 S/15/97
QPTR l 3.2.4 l t ACTIONS (continued) i i CONDITION REQUIRED ACTION COMPLETION TIME l l A;;(pont_iinued) A._4 R_edug.eTPoserJapge Z21hourfifter Neutron 71WHighitrjp each3EllR ($3i2&l84 ! se_tpointsWJf6rW detersiitfMtion K of 0PTR 5 py X 001 M A.45 Reevaluate safety Prior to ! l analyses and confirm increasing l results remain valid THERMAL POWERiand for duration of PoWiF35sie operation under this g rtmen R ux_- condition. lgg!JJj[p @ 2 M' agtEPJ1ptsabove ' thelimQsf 14az-6T Required - ActionsjA.1and 674 M A.56 ---
- NOT @ - - -
, hrd.h.6n A Perform Required
~ @ Action A 56 only after ,
g i,ndea. 4,gn whenevo- 4 c, i, Required Action A.45 is
h J f. ...........
Celibrate cx;;rc Prior to increasing $3;EEn GPTR NdraiaTfzi~erx56rs THERMAL POWER dstector~sN' above the limit [t;1M. __ of Required Action A.1 l ceskore, QMR Vo Hhih lia'i k .j ( Q 3.2. -(.} M (continued) I WCGS-Mark-up ofNUREG-1431 -ITS 3.2 3.2 11 S/15/97
QPTR I ' 3.2.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPi.ETION TIME l l A. (continued) A.67 -- NOTE- - --
- _. _ :-= x .
rK chews kWN Mk' , Ac he i d i A.7 ong =b %"M ' " ed M I R 1r (Ac.t:5nAMb'**P*M
/
Acti A. s in ed.
- j. PerformSR3.2.1.1~,'J 'Jittir, 24 hours
- 3.!2;L 2 and SR 3.2.2.1. after aEhf atun gyg 9!IBU2112E WINM
"--- - i uwd.
gg , hi t$$$18d
.uua.
W, i
@ s.z-(.)
l dn 48 hours
- a er increasing THERMAL POWER above the limit of Required Action A.1 B. Required Action and B.1 Reduce THERHAL POWER to 4 hours associated Completion s 50% RTP.
Time not met. > l l I WCGS-Mark-up ofNUREG-1431 -ITS 3.2 3.2 12 $/13/97
or -
- a i QPTR B 3.2.4 BASES i ACTIONS M With the QPTR exceeding its limit, a power level reduction of 3% RTP for each it by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. T)le Ctap]etionTime is:2 houtsletterleachETR determination;that_lis3erformedf,pjursuant;to R6adired%ctioffE2;."; The Completion Time of 2 hours allows sufficient time to identify the cause and correct the tilt,~MEredQce~powerias n4# M ej. Note that power reduction Q may cause a change in the cilted condition. [qa.z-c.)
~
@ ERT -a -s- i-2.
p 3.2.<,j U After completion of Required Action A.1, the QPTR alarm mey still be in its alarmed state. As such, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours thereafter. If the QPlR continues to increase. THERMAL POWER has to be reduced accordingly. A 12 hour Completion Time is safficient because any additional change in QPTR would be relatively slow.
%m$srium s u.
U to support csinM stoWe. emtMn3 area.chdvwd of h ihtendub when the. co<c w opunhn3 condIho,ns us mappi@ . The peaking factors Pu and Fa(Z) are of primary importance in ensuring that the power distribution remains co1sistent with the initial conditions used in the safety analyses Perfonning SRs on Pa and FM Fa(Z) within the Completion Time of 24j hours ~8Ttbr3chie~ving "M~ '" I
*NMALm equ1 UDr1um.cond1ttonf"^U nf Q;2l:f,2:7= JWTX %^;M PMER reducbb ( fdO. $4 m:,ures Inat these primarf indicators of power distribution
~
P" 89u.sveA Achm, are within their respective limits.# A Completion Time of 24 hours
- --- - t after7schievTnfeqdflibrium ~conditioni takes into consideration the rate at' which peaking factors are likely to change, and the time required to stabilize the plant and perform a flux map. If these peaking factors are not within their limits, the Required Actions-of assocf;ated'tirTth these Surve111ances provide an appropriate response for the abnormal condition. If the QPTR remains above its specified limit, the peaking factor surveillance are required each 7 days thereafter to evaluate Fu' and Fa(Z) with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit.
(continued) WCGS-Mark-up ofNUREG-H31 - Bases 3.2 B 3.2 25 5/15/97
l INSERT B 3.2-25 0 3.2-6 The maximum allowable THERMAL POWER level initially determined by Required Action A.1 may be affected by subsequent determinations of OPTR. Increases in OPTR would require a THERMAL POWER reduction within 2 hours of OPTR determination, if necessary to comply with the decreased maximum allowable l THERMAL POWER level. Decreases in OPTR would allow raising the maximum allowable THERMAL POWER level and increasing THERMAL POWER up to this revised ( limit. l i
\
l i l l 1 i l I
OPTR B 3.2.4 BASES i
<eMm e ACTIONS o 3.2.& - 1 l (continued) )
A reduction of the Power Range . Neutron Flux-High trip,setAointrb.y L3% for each 1% by whichJPTR h exceeds 1:00.~ iEa'/conjLerya2MEtion fot PEotection_against.the_consequetic^es'of severe trans.m , potegially__.unanaly_z.edffpjget d13tritipt10ns.hnceM -- l Required Actionfesults,[n _earT1_er._Qip36tpolmred0ctiM1,d belggui. red. pursuant to the Requirjffctions"d"the FL am specifications. ]he_ Comp]}tio'n~TTine Uf 7Z7f00f5~8Tt~ e r DEEP dttenL111atioq.11..su.fficigilt eqn.]MelibsEg'EfERKt]m severe transient in thi.s time p_eri~ofatWttig33'eceding.pr03.t
~
reduction in ]iERMAL _ POWER ,in ac_cordan,ce Wfth Reaufreid AcIlhDR3 l
%_erasz-}u.4y g 3. 2.<,,, L
\Q3.2.G-1 Although FL and Fo(Z) are of primary importance as initial conditions !
in the safety analyses, other changes in the power distribution may occur as the OPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re evaluation is required to ensure that, before increasing THERMAL POWER to above the limits of Required Actions A.1 and A.4, the react core conditions are consistent with the assumptions in the safety nalyses. mL Power Ry<. HeAm. g3 2 u; I:lu-Hi$ a ig$ctp. A A-5A.6 -- If the OPTR bc; cxcccdcd remains above the 1.02 limit and a re evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are rccclOrcted ic rM QF'r2 shcw a cccc Oi"R normalized toWCAft/SjadtJWt;3B prior to
% JM l.k t.
- > increas1ng THERMAL POWER to above the limit of Required Action A.1.
The process"#fyrmsifzatiun 1F_nlicomp1Tshed'bTmWas~urfrig7Dffiliffts ~fdr each detector during flux mapping and (continued) j WCGS-Mark-up of NUREG-1431 - Bases 3.2 B 3.2-26 S/158 7
l I INSERT B 3.2-26a 0 3.2-6 The Power Range Neutron Flux-High trip setpoint initially determined by Required Action A.4 may be affected by subsequent determinations of OPTR, similar to that of the maximum allowable THERMAL POWER determined by Required Action A.1. I l l 1 i I L_______.________..
QPTR B 3.2.4 BASES
' ACTIONS AJi (continued) using_this informationjoir!ormalize_ttlem D3!p_eacitdsgglar (tittist3tumattga11btat19n of:the N13 otlacough the HEK6 it caTCUTLtt! ggt 10 sutif,.t"mannair"that 12011RREstMM normenzationy.nearTRLShis is done to detect any subsequent significant changes in QPTR.
4 gg pgg Required Action A.56 is _ modified by states that tM ^'" i:; Q until after the e e Miukte, r at ; rad cut 6 regene. qefvt.tu, re evaluation of the sa ety ana ,ys s las ermined that core l sJi+h k Va4 conditions at RTP are wit in the safety analysis assumptions (i.e., Required Action A.45) Thtii Ndt _ intended to prevent any ambiguity about the required e of actions. ga .z-(., } h:,e. Nedin ,
, %.;_z.,4 ---
A-4A.7 ; Once tb '_ % tilt i:: s i;;d ;;t excotE Bulaciars;
,y rubWD
~
j - l 31stewanaAneld- - 13(i.e.. Required Action q:s.z-Q I
@%_M lbb performed), it is acceptab' e to return to full power operation.
However, as an added check that the core power distribution at RTP is consistent with the safety analysis assumptions, Required Action A.6Z l requires verification that Fa(Z) and FL are within their s ified ! limits within 24 hours of.'gchthina;
@sser1as.2-rtQ @3Ff As an added precaution, if the core power does ;
reach RTP within 24 hours, but is increased slowly, then the peaking i
~_ . factorsurveillancesmustbeperformedwithin48 hours 6f'tNe/lBey Mer 3"NW thergseiht J6 36wer'wss painfo These Completion Time _s, are M-N TWAL PowEft- intended to allow T dequate time to increase THERMAL POWER to above the af limit of Required Action A.1, while not permitting the core to remain ph+he h"4 h@M U b with unconfirmed pcwer distributions for extended periods Required Action A.67. is modified by a Note that states that the ,
peaking factor surveillance => crly M de ,; ;'ter sN tern the excore detectors have M- , niitr;;;d t: ;;M s,e tilt n5tmalizedW6'Drffiaf,eWilksteAoCiHTO(1.e. Required QM - Action A.56). "fie intent o'f th1sM;e 1Tto have the peaking factor
-tb (4 & n M surveillance performed at operating power levels, which can only be accom
( * @ plished zfoJ1]gafter er.dthetM excore cer; r;turnddetectors t: ;;r;.;;r.ar; ;;libr;t;d n m to ' [4a.z-c ] (continued) WCGS-Mark-up ofNUREG-H31 - Bases 3.2 B 3.2 27 5/15M
INSERT B 3.2-27a 0 3.2-6 Note 2 states that if Required Action A.6 is performed, then Required Action A.7 shall be performed. Required Action A.6 normalizes the excore detectors to restore OPTR to within limit, which restores compliance with LC0 3.2.4. Tnus, Note 2 prevents exiting the Actions prior to completing flux mapping to verify peaking factors per Required Action A.7. INSERT B 3.2-27b 0 3.2-6 Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions to support flux mapping. i f I \ i
)
DIFFERENCES FROM NUREG-1431 Section 3.2 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431 Revision 1 to make them plant specific and to incorporate generic changes resulting from the Industry /NRC generic change process. ! The change mmbers are referenced directly from the NUREG 1431 mark ups. For ! enclosures 3A, 38, 4, 6A and 68, text in brackets "[ ]" indicates the information is ' plant specific and is not comon to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. 26NGE NLABS JUSTIFICATION , I l 3.2 01 Not applicable to WCGS. See Conversion Comparison Table
- (Enclosure 68) 3.2 02 Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 68) l 3.2 03 Consistent with TSTF 98, Rev.1 the factor by which the Fa must I be adjusted on increasing Fa measurements is moved to the COLR. i This change is acceptable because thes factor is normally l l
contained in the COLR, and it removes detail not required to be i contained in TS. i l 3.2 04 Not applicable to WCGS. See Conversion Comparison Table ; (Enclosure 6B) l lus5RT di A- L_ _ 3.2 05 ('Pera sion 95 _ cha arifi t
. ors real to 1 e zero . T is ly _ clarif tion of G 1431 rding a. is acc able [qn-s,[
3.2 06 Consistent with TSTF 95, the time allowed for resetting the power range neutron flux - high setpoint if Fo or N is outside their limits is extended from 8 hours to 72 hours. As written, the Completion Time of 8 hours to reduce the Power Range Neutron Flux High trip setpoints presents an unjustified burden on the operation of the plant. A Completion Time of 72 hours will allow time to perform a second flux map to confirm the results, or, detemine that the condition was temporary, without implementing an unnecessary trip setpoint change, during which there is ! increased potential for a plant transient and human error. l Following a significant power reduction, at least 24 hours are required to re establish steady state xenon prior to taking a flux map, and approximately 8 to 12 hours to obtain a flux map, and analyze the data. A significant potential for human error can be created through requiring the trip setpoints to be reduced within the same time frame that a unit power reduction is taking WCGS-Differencesfrom NUREG-1431 - ITS 3.2 1 S/1537
l l l i INSERT 6A-1 0 3.2-6 ! 3.2-05 Consistent with TSTF-241 ISTS 3.2.4, Guadrant Power Tilt Ratio, is revised to provide more appropriate Actions. Required Action A.2 contains a redundant action to reduce THERMAL F0WER. This redundant action is deleted and the THERMAL POWER limit of Required Action A.1 is revised to provide the appropriate allowance for subsequent power reductions based on subsequent determination of OPTR. [This was approved in CTS in Amendment No.116 for Wolf Creek.] The Completion Time of Required Action A.3 requires the peaking f actors to be verified within 24 hours of achieving equilibrium conditions with THERMAL POWER reduced by Required Action A.1. In the current Required Action, a significant fraction of the 24 hours could be spent waiting for the plant to stabilize at the new power level leaving insufficient time to measure and analyze the peaking factors or resulting in the peaking factors being measured when the t plant is not stable yielding inaccurate information. Since the peaking factors are of prime importance, the proposed change will allow sufficient time to obtain an accurate measurement. [This was approved in CTS in Amendment No.116 for Wolf Creek.] Required Action [A.6] is revised to add a new Note stating " Required Action [A.7] shall be completed if Required Action [A.6] is performed." As discussed in Section 1.3 of the ITS, an Actions Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not with the LCO Applicability. Therefore, when Required Action'[A.6] is completed, OPTR should be back within limit and the LC0 may be exited. Adding this Note ensures that the peaking factors are verified after normalization of the excore detectors. Additionally, Required Action [A.6] is revised to state " Normalize excore detectors to restore OPTR to within limit." Normalization is accomplished in such a manner that the indicated OPTR following normalization is near 1.00. Thus, the absence of a tilt will manifest itself as OPTR-1.00 rather than zero since quadrant power tilt is expressed as a ratio. Also, from a literal compliance standpoint, the tilt cannot be restored to exactly 1.00. [This was approved in CTS in Amendment No.116 for Wolf Creek.] _______ _______ _ U
CHANGE NUMBER JUSTIFICATION e aype d in b d M M2M l A)o.I16. ' 7 n t [ i 3.2 18 This change modifies the requirements in NUREG 1431. Rev.1 l for_ Wolf _ Creek _to retain current TS requirement l l (1 e pdvigony fo tiequipea Acyrops_"pr9posperD7 l ( -
.fThe current T5 provisions that would be retained are (1) the requirement to reduce the Power Range Neutron Flux High trip 1 l
setpoint during a required power reduction, and (2) the I restriction that incore flux mapping can be used to verify QPTR l only after one (not "one or more") excore neutron flux channel is 1 inoperable with power greater than 75%. Retention of the current reactor trip setpoint reduction would add a new Required Action to the QPTR specification. However, the Completion Time would be revised from 6 hours to 72 hours l t based on industry Traveler TSTF-95. These changes are acceptable l because resetting the trip setpoints for QPTR not within limit is ; l based on an assumption that peaking factors are not within limit. ! The Required Actions of NUREG 1431 also would result in resetting the trip setpoints but only after Fa and F% measurements have I been taken and determined to be not within limits. Therefore, the proposed change is acceptable because it would result in resetting the trip setpoints at an earlier time in the l performance of the Required Actions. The extension of the time
- period to reduce the flux trip setpoints (from 6 to 72 hours) is l also acceptable because it would permit time to perform required
- QPTR determination, permit orderly resetting of the high flux l trip setpoints, and reduce the chances of an inadvertent reactor trip during the required power reduction. The extension of time l to 72 hours is consistent with industry traveler TSTF 95.
l Retention of the incore flux mapping restriction would delete the ; words "or more" from the Note to SR 3.2.4.2. This change is acceptable because it would prevent operation with less than three excore power range channels available for QPTR measurement l and comparison to the incore flux mapping results. This would provide additional, conservative margin in using the incore detectors to verify QPTR and additional capability for monitoring QPTR between incore maps. g 3.r..c.\ The c bas o acceptabl cause the allow a onal ti r the nt conditi to stabil be e ifying ing fa rs. The ired Actio of G 1431 would provi fficient for the nt t tabili I y large r reductio s to obtai n accu e measu ntoff WCGS-Differencesfrom NUREG-1431 - ITS 3.2 4 S/15/97
l I l CHANGE NUMBER JUSTIFICATION y G 3.2-4. l
, <~ '
,. A 95 o rifi the R 1r ction y ma 1
i ear (fiat a r ucti is r red thin hours ach da _ in on. addit n, t cha ext s the 7 l l#M695 ept achiev g equ 1bri plan ondi ns to i pea fact measur nts e uir to rfo dur any ppoweri ases f owingpderr ctiot(srequredby QPTR; la l 3.2-19 hisWo5f eek s liter camp ec ific mod ies SR 3 rn regar ng the
.3.1 t resolve) of t CO Note o
, yt . Note st s that or mor xcore c nels i icat out of mits be e the L is not must . I l ver. 3.0.1 st s that ure to t a SR s 1 be l jfail to meet LCO. F ure to SR 3.2.3 in IST l d occur i nly one core cha were i rable. nce the pote confli between t te and SR. s cha is ac able se it maint s the i nt of the t TS the I for AFD whi eliminati a lit al comp 1 nce bfk q3.Een}
l l WCGS-Differencesfrom NUREG-1431 - ITS 3.2 5 5/158 7
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-7 APPLICABILITY: -WC, CA REQUEST: ITS 3.2.1 Heat Flux Hot Channel Factor CTS 3/4.2.2 Heat Flux Hot Channel Factor (Callaway & Wolf Creek) JFD 3.2-17 ITS SR 3.2.1.1 Frequency Comment: The ITS SR 3.2.1.1 Frequency does not adopt the STS of "within [12] hours" based upon the justification that the CTS does not specify a time limit. The STS uses a l bracketed time for accomplishing the SR, meaning that a plant specific number can be utilized. Utilize a plant specific number based upon plant experience, or other relevant , justification, if the 12 hours is unrealistic. A time limit must appear in the frequency in l place of the brackets. l FLOG RESPONSE: Callaway and Wolf Creek are placing a time limit in the Frequency for ITS SR 3.2.1.1 and SR 3.2.1.2 in lieu of maintaining CTS which did not specify any completion time. DOC 02-13-LG (applicable to Cal!away only) and JFD 3.2-17 are no longer used. See response to Comment Number 3.2-3. ATTACHED PAGES: l See attached pages for response to Comment Number 3.2-3. I l l o
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.2-8 APPLICABILITY: WC REQUEST: ITS 3.2.4 Quadrant Power Tilt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (Wolf Creek) JFD 3.2-18 DOC 04-10-LS14 Revised ITS based upon CTS
. Ccmment: JFD 3.2-18 justifies numerous changes to the STS based upon CTS requirements, some of which are unacceptable. The unacceptable STS changes are:
- 1. The editorial change to the Required Action A.1 Completion Time and the associated change to Required Action A.2.
- 2. The qualification of "after achieving equilibrium conditions . " in Required Action A.3 Completion Time.
- 3. The qualification of "after achieving equilibrium conditions . " in Required Action A.7 Completion Time.
In each of the above instances, provide adequate justification for the change or adopt the STS version of the specification. ! FLOG RESPONSE: The proposed changes discussed in JFD 3.2-18 were approved for Wolf Creek in Amendment No.116. Additionally, several of the changes discussed in JFD 3.2-18 are proposed in TSTF-241. JFD 3.2-18 and JFD 3.2-05 were modified to differentiate between the changes associated
- with TSTF-241 and those changes approved by Amendment No.116.
The changes discussed in JFD 3.2-05 were also approved for Wolf Creek in Amendment No.116. j ATTACHED PAGES: See attached pages for response to Comment Number 3.2-6. I i i k
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-9 APPLICABILITY: WC
'XQUEST: ITS 3.2.3 Axial Flux Difference CTS 3/4.2.1 Axial Flux Difference (Wolf Creek)
JFD 3.2-19 ITS SR 3.2.3.1 Comment: The ITS deletes the STS words 'for each OPERABLE excore channel," based upon a literal compliance concern. The concern is adequately addressed by the Note following the LCO, and is a generic change. The words are included to ensure that the AFD is verified with all the excores. Submit a TSTF change request providing adequate justification for the proposed change, or adopt the STS version of the SR. FLOG RESPONSE: Wolf Creek proposed the above change based on a perceived literal compliance concern within this specification. The SR 3.0.1 Bases states that SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. In this case, the Note is not associated with the SR. Wolf Creek understands the logic behind this specification to be that if AFD is not within limits on one OPERABLE excore channel, the LCO would be met because the Note on the LCO indicates that the AFD is only considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits. Wolf Creek is proposing to revise the ITS SR 3.2.3.1 Bases to state: "Two or more OPERABLE excore channels outside the AFD limits would constitute failure of this surveillance." Revising the ITS SR 3.2.3.1 Bases clarifies that having one excore channel inoperable does not constitute a failure of the surveillance requirement and subsequent failure to meet the LCO. This Bases wording is consistent with CTS SR 4.2.1.2 which states that: "The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits. Wolf Creek requests that the staff document in the Safety Evaluation Report (SER) the above position. With this position documented in the SER, Wolf Creek withdraws the proposed change to ITS SR 3.2.3.1. l ATTACHED PAGES: Encl. 5A 3.2-9 Encl. SB B 3.2-21 Encl.6A 5 Encl. 6B 3
AFD (RAOC Methodology)- 3.2.3B 3.2 POWER DISTRIBUTION LIMITS 3.2.38 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) LCO 3.2.3 The AFD in
- flux difference units shall be maintained within the limits specified in the COLR.
............................N0TE -- -- - -- ~- ' ----- --
The AFD shall be considered outside 11Mts when two or more OPERABLE excore channels indicate AFD to be outside limits. I APPLICABILITY: MODE 1 with THERMAL POWER a 50% RTP. i l ACTIONS COWITION- REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER to 30 minutes ;
< 50% RTP, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY hs.t-il Verify AFD within limiteIfer ;;;t. 0""'AC SR 3.2.3.1 __ _-
7 days
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WCGS. Mark-ap ofNUREG-1431 -ITS3.2 3.2 9 5/158 7
AFD (MOC Methodology) B 3.2.3 BASES (continued) ACTIONS 61 As an alternative to restoring the AFD to within its specified limits, I Required Action A.1 requires a THERMAL POWER reduction to < 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating ) experience, to reach 50% RTP without challenging plant systems. SURVEILLANCE SR 3.2.3.1 I REQUIREMENTS
" A~ i; ..witer;d en
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This Surveillance verifies that the AFD n ir.dist;d by the M0 sere cherel, is within its specified limits ad i; -...;i:;ter.:. with th; :;teta; ef th; ^~ ..~r. iter ;l- ;. MRIMsN I agh951ABLE,jMISlemichanneliatteeTangcjjeI,ggglLcjg _ , _ _ m m_. aWi&4he _
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s iter ;1e.; 0"C" C ". t The Surveillance Frequency of 7 da."s is adequate considering that the AFD is monitored by a computer and any deviation from requirements is alaned. Tw. m. . oreRAe.t.E encue chMS m hide h App iiW;b wcuta c,n.Mhd< 42% of 4W.5 s u.u<.i ll w. l (continued) ) WCGS-Mark-up ofNUREG-H31 - Bases 3.2 B 3.2 21 SR587
CHANGE NUMBER JUSTIFICATION q Q 3.2 -(,
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, yt . Note st s that or so xcore nels ' I must icat out of mits be e the L is not .
i i ver. 3.0.1 st s that ure to t a SR s 1 be ! jfail to meet LCO. F ure to SR 3.2.3 in IST l i d occur i nly one core cha were i rable, e I i the poten confli between t te and SR. s cha is ac table se it maint s the i nt of the nt l TS the I for AFD whi eliminati a lit al compl nce b#h Q 3. E-9} l WCGS-Differencesfrom NUREG-H31 - IIS 3.2 5 5/158 7 1
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.2-10 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.2.4 Quadrant Power Tilt Ratio CTS 3/4.2.4 Quadrant Power Tilt Ratio (All FLOG Plants) JFD 3.2-15 ITS SR 3.2.4.2 Comment: JFD 3.2-15 justifies numerous changes to the STS one of which is unacceptable. JFD 3.2-15 is based upon TSTF-109, which has been rejected. The unacceptable STS change is: The modification of the note to SR 3.2.4.2, and in particular the addition of the 12 hour allowance in the Note to SR 3.2.4.2. Provide adequate justification for this change or adopt the STS version of the Note. FLOG RESPONSE: The latest status report from the TSTF industry database, dated June 16, 1998, indicates that the NRC has approved TSTF-109. The FLOG continues to pursue the changes approved in TSTF-109. JFD 3.2-15 is revised to delete the sentence: "The note for SR 3.2.4.2 it. changed to require performance if one 'or more' QPTR inputs aie inoperable." and added: "The note and Frequency for SR 3.2.4.2 are revised consistent with typical presentation formats that provide for a l period of time after establishing conditions." NUREG-1431, Rev.1
]
currently has "or more" in the Note and TSTF-109 did not modify this wording. ATTACHED PAGES: Encl.6A 3 Encl. 6B 2 i
['Thz 4. Frqueue[b SU.2.4.2. are raNded
' con >Me4 w% tyP's cM presentation &rmd.sihat provide Viie a pahood of -r:4c. af CL, estatA!.isQ c.,nAihens '.JijSTIFICATION ~-
CHANGE NitBER -- 3.2 12 bt a c e
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3.2 13
, a s. 3. -d This change retains the CTS for the performance or peaking factor determinations following plant shutdowns. The CTS, through the exemption to specification 4.0.4. allows prerequisite plant conditions to be obta1ned prior to requiring that the surveillance be completed. { sex [r,A-%RQ 3.2 4 3.2 14 This change retains the Wolf Creek CTS for the completion time for Required Action 3.2.2 A.2. This Completion Time was approved ,
in License Amendment 61. This change is based on the time l required to reduce power establish equ111brita conditions, and ' obtain a flux map. 3.2 15 This change incorporates industry traveler TSTF 109. Action A.2 would require the QPTR be determined rather than performing a }
. specific surveillance because more than one surveillance can be used to determine QPTR. SR 3.2.4.1 was revised to retain allowa e that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1.)
. . .. ":r ". ;'.:. 2 W t 1 te w re_ u. k. . a n 6- Q
_ _ . _ ,. . . ,_ . _. .. . .;f These c1anges are L; 4.5.2.-10 { acceptable because they clarify the ISTS regarding frequency and use of incore flux monitoring for QPTR measurement. The changes reflect that incore detectors provide an acceptable QPTR determination during all plant conditions. 3.2 16 This change would require that both transient and static Fa measurements be determined when performed for Required Actions 3.2.4 A.3 and 3.2.4 A.6. The intent of the Required Actions is to verify that is within its limit. Fa(Z) is approximated by F*a(Z) (which is obtained via SR 3.2.1.1) and F"a(Z) (which is obtained via SR 3.2.1.2). Thus, both Fen (Z) and F",(Z) must be established to verify Fa(Z). This change is consistent with traveler WOG 105. 3.2 17 Mr ncy r remen or perfo ng FqMasur s revi to onn to which not specify a let nT . pr tice i o perfo measurement as as racti . The SR Coup ion Times are sed o what sa no ly rea le Coup ion Time for formi a fl map: / ver, i problems cur, the plant be f ed t reduc ' power o shutdown. iswould'subj the nt to tran ent cond on wit suffici safe basis There re, ma' taining current requ ement s accep ble ause i
@ useA. / kos.2-Sl WCGS-Differencesfrom NUitEG-1431 - ITS 3.2 3 5/15M7
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ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.2-001 APPLICABILITY: DC, WC, CA REQUEST: Revise last sentence of Bases 3.2.1, Action B.1 to read: " Reducing both the , positive and negative AFD limits by . . " This change makes it clear that both l positive and negative limits must be reduced when Fo*(z) is not within limits. ATTACHED PAGES: Encl. 5B B 3.2-6 l l I i t
4 i Fe(Z) (Fe Methodology) B 3.2.1 I BASES ACTIONS ]L1 (continued) 1 If it is found that the maxinam calculated value of Fe(Z) that can occur during normal maneuvers, $Z), exceeds its specified limits, there exists a potential for $Z) to become excessively high if a ) normal operational transient occurs,(Se6uef'pgihF#$by a it for j each it by which $Z) exceeds its Visit within the allowed Completion ' Time of-4 4 hours, restricts the a) 1al flux distribution such that even if a transient occurred, core peaking factors are not exceeded. N 'g % s n e y b 4 h t h e yibn y LJa s.2.- W **'g' ## # "j' C43.1aos[ If Required Actions A.1 through A.4 or B.1 are not met within their ! associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours. , This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from j full power operation in an orderly manner and without challenging plant systems. 4 SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The REQUIREEKTS Note applies irir; tt.; fi .;t ;;;ar enesim .ft r ; refalir; during , powercascensionsj@lowigtajplantishutdouniRea_v1M MysDhe 1 note alfous;fDrfpower c asednsionsliFtheisalgillancescamputacurmit. It states that THERNAL POWER may be increased until an equilibrium l power leve1@e5.Jegu1Hbntissiconditions) has been achieved at which ; a power distribution map can be obtained. , This allowance is modified, however, by one of the Frequency conditions that requires verification that $Z) and $Z) are within their specified limits after a power ; rise of more than 10% RTP over the TERMAL POWER at which they were last verified to be within specified limits. Because $Z) and $Z) i could not have previously been measured in-4Ms a reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of $Z) and $Z) are made at a lower power level at which adequate margin is available before going to 100* RTP. Also, this Frequency condition, together with the frequency condition requiring verification (continued) WCGS-Mark-ap ofNUREG-H31 - Beset 3.2 8 3.2 6 5/1537
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.2-002 APPLICABILITY: DC, WC, CA l REQUEST: Insert in ITS Bases 3.2.1 the expressions from CTS SR 4.2.2 for how to calculate ' the percent by which both Foc and Fo*' exceed their limits. This provides more readily available information to Operations and Engineering personnel. ATTACHED PAGES: Encl. SB B 3.2-5, B 3.2-6 l 1 l l l l l
Fa(Z) (Fo Methodology) B 3.2.1 BASES-ACTIONS U (continued) plant to remain in an unacceptable condition for an extended period of time. hsautT S 3.2-F) cA 3. 2. -oc) 2- } u A reduction of the Power Range Neutron Flux-High trip setpoints by
- t 1% for each 1% by which F8(Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of-8 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.
O Reduction in the Overpower aT trip setpoints by :t 1% for each 1% by which F{(Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.I. l Verification that F{(Z) has been restored to within its limit, by l performing SR 3.2.1.1 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels are consistent with safety analyses asstaptions. 7thertnt~TrQl hts 7attTon minedtifii:stan oT the cliiBe s Ice oGt~dfmKcdH5ftYeinWM__Mtih JE3fi~e cSiteiTo"the[egtent .r!acessarKtelaTla. wTsJf03lilitq1gn;4phr power.Jeweli (continued) ! WCGS-Mark-up ofNUREG-1431 - Bases 3.2 B 3.2 5 S/158 7
INSERT B 3.2-5 CA 3.2-002 Calculate the percent Fa (Z) exceeds its limit by the following expression: C r - - 3 ) maximum F[(Z) -1 - X 100 for P 2: 0.5 over Z CFQ
, X K(Z)
A _ _
/
r - - 3 maximum F[(Z) -1 - X 100 for P < 0.5 over Z CFQ 05 X K(Z) . l l
Fa(Z) (Fa Methodology) B 3.2.1 BASES ACTIONS IL1 (continued) l If it is found that the maximum calculated value of Fa(Z) that can occur during normal maneuvers. Q(Z), exceeds its specified limits, there exists a potentia? for Q(Z) to become excessively high if a normal operational transient occurs (.Se6uefbfW!@)by 2 it for each it by which 4(Z) exceeds its mit within the allowed Completion Time of-B 4 hours, restricts the a) al flux distribution such that even if a transient occurred, core peaking factors are not exceeded. in U 'Ia 2.2.-@ rkM.n.b,.b.4h
, , , ih2 4,= pse
, g,g If Required Actions A.1 through A.4 or B.1 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours.
This allowed Corpletion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The REQUIREMENTS Note applies iring th: first p;;. s;;nsion ;ft r ; r;faling dur;ing powegascerisionstf6110 wing;a plantishutdowit(reaving:NogM~The note;allWsEfor powergasednsionsif 'thefsurv_ef1Tancesra&Mcurrent. It states that THERMAL POWER may be increased until an equilibrita power level.gi;egequilibriumscondjtions) has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that $Z) and Q(Z) are within their specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which they were last verified to be within specified limits. Because $Z) and Q(Z) could not have previously been measured in-th+s a reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of Q(Z) and Q(Z) are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification (continued) WCGS-Mark-up ofNUREG-1431 - Bra s 3.2 B 3.2 6 5/1587
l i INSERT B 3.2-6
- Calculate the percent F/(Z) exceeds its limit by the following expression:
-3 maximum F[r; X W(Z)
~
-1 X 100 for P 2 0.5 overZ CFQ P
X K(2) 3 maximum Fj(Z) X W(Z)
~
-1 X 100 for P < 0,5 over Z CFQ X g(7) 05 .
l l I
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CP 3.2-001 APPLICABILITY: DC, CP, WC, CA ! REQUEST: The mark-up CTS SR 4.2.5.3 (4.2.3.4 for DCPP) was revised as follows. This SR requires a CHANNEL CALIBRATION on the RCS loop flow rate l once per 18 months (refueling interval for DCPP). The CTS SR is l equivalent to ITS SR 3.3.1.10 and the Reactor Coolant Flow - Low I functional unit. DOC 5-12-A was initiated to address that CTS SR 4.2.5.3 l ( .. 3.4 for DCPP) is equivalent to ITS SR 3.3.1.10. For CPSES and l DCPP, the strikeout is removed consistent with the FLOG markup methodology. For CPSES, the CTS SR 4.2.5.3 statement "The channels shall be normalized based on the RCS flow rate determination of Surveillance Requirement."is struck through and DOC 5-03-LG applied. DOC 5 LG is revised in Enclosures 3A and 3B to indicate the DOC is applicable to CPSES only and that this information is moved to ITS Bases 3.4.1. ATTACHED PAGES: Encl. 2 2-15 Encl. 3A 8, 9 Encl. 38 7, 8 I
)
l
)
l l
POWER DISTRIBtJTION LIMITS 3/4 2.5 DNB PARAMETERS l Li'.tiTING CONDITION FOR OPERATION i ACTION: (Continued
- 4. ! der $j nd hrrect the -"r Of the out-of-limit condition !!fs$0$IlliF prior to incressmg THERMAL POWER above the reduced THERMAL POWER **h*EN limit required by ACTION 1.b and/or 3, above; subsequent
- POWER OPERATION may proceed provided that the indicated RCS total flow rate is demonstrated to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels
- a. A nominal 50% of RATED THERMAL POWER,
- b. A nominal 75% of RATED THERMAL POWER, and
- c. Within 24 hours of attaining greater than or equal to 95%
of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.5.1 The provisions of Specification 4.0.4 are not applicable to Specification. 3.2.5.c. 4.2.5.2 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. ' DF-12. -A 4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL I CP 3.2-00 Il Y CALIBRATION at least once per 18 monthe. l J
-- .2 4.2.5 A The RCS total flow rate shall beineasured l 1:'- .:d bj prM9^- hret balam.. awasurement at least once per 18 months @ '.^"- ' d:/:
;Mer g&gg.g7q Q nahm
. . _ a te m ':-
50-t".5_
; 5_ _;n_ .^'_-'r-
_ _ _, _. _ __ _ _ ___m.._,__..__..,_ _ . . ,_ ,n , N :x'. ;m ::" :f-- , _ _ . . _ _ ,
'E.Y NN 5 blN 5 "$ 5 Y'~- E:NtEn 8 5 5 ' Y.E*]ibEE? En: ~ ((MIAi?
manicssass 4.2.5.5 5 '-*feter ;:rted th !! 5 !reperd fer '-"" ;; nd W:::d 2:
--- r ; et 'r * - = p r 1 ? : ' .:.
]
{# Not required to be performed until 7 days after achieving 2 95 % RTP.) FIO5474I" mumm WOLF CREEK - UNIT 1 3/4 2-15 Amendment No. 61 Mark-up of CTS 3M.2 S/13/97 _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ ~
CHANGE j NUPBER gile DESCRIPTION m.g, l
- n. roposed hanges, bas' n traveler 95, flimina whi clarif t reductions be comp 1 withi 2 l
rs a each determina and permi chievi l ,equ rium c 1tions for suring peak factor re npt ' l L nsider o be relax ns of curre equir s. A s is l beca completion s for thes ctivitie re not , l = c -- ~ specified in cu ent TS.f l Cod -11 A Nssett an .e>j I we. s. a. -oo a 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 l detail out of the specification is consistent with NUREG 1431 and is an example of removing unnecessary details from the TS in accordance with 10 CFR 50.36. 05 02 LS-7 Not applit "le to WCGS. See Conversion Comparison Table l (Enclosure 1). gcP 3.7.-oo d 05 03 LG l [ sten I ith ION EG 14 ,thepequit he RC5 f1 t me rs t 1 st nce fo a 18 hs the uirement't no 11 t c 1 ae i ed the sfortgRCffl - r cto tr f on in Sectior(3.34. Not o esosconver:s'en chep.m..im rwe C.ppuc2LAA etnea. m w w a intM/sS-) s L ,/ 05 04 LG Consistent with industry traveler TSTF 105, the explicit requirements that the RCS flow be measured through the use of a precision heat balance measurement and that the instrumentation used in the perfomance of the calorimetric flow measurement be calibrated within a specified time period of perfoming the measurement is moved to a licensee controlled document. The requirement to verify that the RCS flow is within limits remains within the Technical Specification. This is an example of removing unnecessary details from the TS and is acceptable based on the guidance provided in 10 CFR 50.36. 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 l required to be in TS to provide operational safety while
! retaining the limiting conditions for operation. l 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 i parameters of Reactor Coolant System (RCS) average j temperature pressurizer pressure, and RCS flow rate are l WCGS-Description of Changes to CTS 3N.2 8 5/15/97 i
CHANGE , Nul6ER EiHC DESCRIPTION maintained within specified limits in order to ensure consistency with the asstmed initial conditions of the accident analyses. The limits placed on the RCS temperature, 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 requirea plant conditions in an orderly manner without unnecessarily initiating any undue plant transients and on the small likelihood of a severe event occurring during the extended time period. l 05 07 M This surveillance for measuring RCS flow by precision heat balance is modified to add a footnote that corresponds to the Note for ITS SR 3.4.1.4. The footnote requires that the i surveillance be performed within 7 days of achieving 95% RTP. l This is more restrictive in that it ties the surveillance to l the beginning of a cycle. This is acceptable because other l indication of RCS flow is available (RCS flow meters) and time I is provided to establish plant conditions suitable for the precision heat balance. This is consistent with traveler WOG-
- 99. In addition, the THERMAL POWER specified in the footnote would be changed from the generic value provided in NUREG 1431 to a plant specific value of it 95
- RTP. This change is acceptable because it specifies a THERMAL POWER in better agreement with current operating procedures for performing a precision heat balance. Current TS do not 'specify a power level for this measurement.
l 05 08 Not used. 05 09 LG The requirements for inspecting and cleaning the feedwater flow venturi would be moved to licensee controlled documents.. These details are not contained in NUREG 1431. This is an example of moving unnecessary detailed information from the TS , and is acceptable. l. 05 10 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 05 11 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38).
- 2. A iM5EltT 3A . .S] } GP 3."2. -00 i \
- WCGS-Description of Changes to CTS 3M.2 9 5/158 7
INSERT 3A-9 CP 3.2-001 5-12 A The requirement to perform [18] month CHANNEL CALIBRATIONS of the RCS loop flowrate indicators is part of ITS SR 3.3.1.10 for Reactor Trip System Instrumentation Function 10 (Reactor Coolant Flow - Low). i I l-l i
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T R R E 2 E B 1 _ S M - N U 5 I N 0A
ADDITIONAL INFORMATION COVER SHEET ( ADDITIONAL INFORMAllON NO: TR 3.2-004 APPLICABILITY: DC, CP, WC, CA REQUEST: Revise Traveler Status page to reflect NRC approval and latest revision number of travelers TSTF-110 Rev. 2 and TSTF-136. There are no changes involved to any CTS mark-ups, ITS mark-ups, DOCS, or JFDs. ATTACHED PAGES: Encl. 5A Traveler Status page i I 1 l l
--____________A
Industry Travelers Applicable to Section 3.2 TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-24 Not Incorporated NA Not NRC approved as of traveler cutoff date. TSTF-95 Incorporated 3.2-06 Approved by NRC. l TSTF-97 Incorporated 3.2-07 Approved by NRC. TSTF-98, Rev.1 Incorporated 3.2-03 TSTF-99 Incorporated 3.2-08 Approved by NRC. TSTF-109 Incorporated 3.2-15 Approved by NRC. TSTF-110, Rev.(h Incorporated 3.2-10 @ M @ T r. 3.t a04) TSTF-112, Rev.1 Not Incorporated NA Not NRC approved as of traveler cutoff date. TSTF-136 Incorporated NA [ppwe S.d0o4) TSTF-164 Incorporated 3.2-11 Applicable to CAOC only (CPSES). _
'$6D5, Yorporated 3 E ,ihB48- Willincorporate portions I
( l y ;- y" f.2 __ s . 2.- o9 ofTSTF-25. 1 WOG-105 Incorporated 3.2-16 J e
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.2-001 APPLICABILITY: WC, CP REQUEST: CTS SR 4.2.4.1 and 4.2.4.2 markups for Wolf Creek did not include ITS SR l 3.2.4.1 Note 2 and SR 3.2.4.2 Note. Adding these Notes to the CTS markup l resulted in creating DOC 04-11-A which is only applicable to Wolf Creek is maintaining the CTS requirement for performing SR 4.2.4.2 with only one Power l Range channel inoperable and not adopting the changes proposed in DOC 04-04 LS-12. For CPSES, CTS SR 4.2.4.2 was not marked up to include ITS SR 3.2.4.1 Note
- 2. The Note has been added. The addition of the Note is covered by existing DOC 04-04-LS-12.
ATTACHED PAGES: Encl. 2 2-12 Encl. 3 7, 8 Encl. 3B 6, 7 Encl. 58 B 3.2-28 l 1 l l
I45ERT 2-11 Q 3.~2.-( l POWER DISTRIBUTION LIMITS I-LIMITING CONDITION FOR OPERATION ACTION (Continued)
- 7. Within 24 hours after achieving equilibrium conditions not to exceed 48 hours after increasing THERMAL POWER above the limit i of ACTION a.1.**- i a) Confirm that the Heat Flux Hot Channel Factor F (Z) is within its limit by performing Surveillance Req,uirement 4.2.2.2. and l b) Confirm that Nuclear Enthalpy Rise Hot Channel Factor.
F%.iswithinitslimitbyperformingSurveillance Requirement 4.2.3.1.
- b. If the requirements of a.1. , a.2. . a.3. . a.4. . a.5. . a.6. . or a.7.
above are not met, reduce THERMAL POWER to s 50% of RTP within the I next 4 hours.
- c. The prevision: Of Speci'4 cation 3.0.4 cre not ;pplicable. 4-ot-A l
SURVEILLANCE REQUIREMENTS . 4.2.4.1 The OVADRANT POWER TILT RATIO sha ll be determined to be within the DIM. limit above 50% of RATED THERMAL POWER by:lt) 4-M-AT 4-in-AJ
- a. Calculating the ratio at least once per 7 days ahen the alor: i; oi-o1-u;r-
-OPEP,AOL:. onu
^ Calculating th: retie et le65L ence cr oper:tionwhentheci;r:i;in;;;r:bTc.12hoursduiing5te6dy-5t6te 8 8*-t*
4.2.4.2 The OUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors t: : nfirr. th;t the 4 63-uc tric normali:0d sy=:tric th;;b g = :le le.p;w;r 6tiona vidistribution. a full sui e Obt:in;d ilua iii6p.frc; i; ceasi5 twc ;;ts tentcf f;;r ith the 4edicated 00^0P.^NT POWEP, TILT RATIG at least once per 12 hours (+{ 4-in-Ay w.s.2..oail
+ 14% kpat -bm *ne. Ader Range. t4.uba Flu charmel ineptmMe. 4-C4-A anA.THannAt potaew. 415% RTP, 4he. rernainin,9 hue. peuner on4f-channels ca.n be. ux4 for c tcam ong..
ES Mil {SR. 4.2 4 2. m3be. pev-ftwvned in isi of M ew ve}hma.. 4-il-A l _**ACTION a.7. must be completed when ACTION a.6. is performed. t -G m o n e. it_4
-++ seAVequwed + be. perf cene<L ue.l if hour afNf cng%
Power Eenhe. e4eu.tv n h channet to A peraw. w ntainAt-.
- 3 *ll
-n . are. , - -
WOLF CREEK - UNIT 1 3/4 2-12 Amendment No.116 (Next page is 3/4 2-14)
CHANGE litEEE JGE DESCRIPTION 4 3.7. (, { to be i ed y educ g he es blis ng grpater ma t f ign a s, t r tion f r ui t IT r it na nt t nsi that ra uld s ere han red tion o le th
- r uir CT fur r, e mina gt tri et nt r u on ept e ause QPTR n exc so imi p ssa y1 yt acci nt a yse ass tio I av vio ed: (3) IT Requi A ion ri o s seg t to r ensio prov as ran th ppver
)o atio at or ar R will in cord ce wi h the sdfetyj ly and, heref e. ac tabl g w s 04 07 A The statement that Specification 3.0.4 does not apply is no longer needed as revised Actions permit continued operation for an unlimited period of time. In accordance with the provisions of ITS LCO 3.0.4, this is acceptable.
04 08 Not used. 04 09 A St ;;H:ldi te N. k Cvi,mi oivu CuiiymdvM uc M-a00 {Ed:re % iWmeT 3A- 7j - M used .T hM 04 10 The r at cha ) wo y e on r ett t s Ra r ise the on F1 82M !
\ !
! ( ri set n dur t r ion ollow QPTR ; ! su , dc ify t rr cti aus lc et w its r aft ea TR terni tion a evi eq ibri c ition for suri aking f tor , a (3) id imi te t curr TS S r ri QP o est ed wi in 2 urs, to ri d ng etu to r, rese ng the r
,R N ron ux igh 1p int < 55% lowi powe ed on 50% or low. urrent requi a rr cti wit 2 s of terni of limit and igh ux 1p se int uctio ithin next j rs. leti time or res ing t high trip ,
set ts uld cha to 7 urs er de ing TR out lim Th is ac tabl ause Id rait 1 ti o pe rm r red de ination], permiorderly f i setti of t igh ux tr setpoipts, and uce
)cha s of 'inad ent actor trip duri the r 1 red rr tio Eli ting t/curre S ACT S requ ing jQP be tor ithin 244murs, to verifi
/ ing etur opower,,4ndres ing t ower R I Neu n F1 High rip set to < foll g r ir uctio o 50%
or is ac table; aus , they are. cons ent wit EG- 1. Rev and, a te in N LS- 4, the reg ' ements ld pro acc able alternatives to CTS Actions that are pro _ sed for WCGS-Description of Changes to CTS 3N.2 7 5/158 7 t
r __. ___ _ _ INSERT 3A-7 WC 3.2-001 04-09 A Consistent with NUREG-1431, Rev. 1, a Note is added to permit three OPERABLE excore channels to be used to calculate OPTR when one channel is inoperable and power is s 75% RTP. This is consistent with operations permitted by the combination of the CTS definition of OPTR and the CTS surveillance requirement for verifying GPTR by use of the incore detectors when an excore channel is inoperable. 1 l l l
CHANGE NUEER EC DESCRIPTION g 3,2 5 } 1 na 'n. roposed hanges, ba I n traveler whi clarif t reductions be compl . withi 2 rs a each determina and perai chievi lequ rium c tions for suring peak factor re npt ' L nsider o be relax ns of curre equir s.gs is
!beca completion s for thes ctivitie re not
=- ~ specified in ca ent TS.f 6M -11 A iesser sA e%1 we. s. a. -oo i 1 l 05 01 LG The designation of how instrument uncertainties are treated l (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 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). fcP 3.L-ooI \ l 05 03 LG [pefsten ith EG 14 ,thepequif fo ' a l GRANNE I ION he R fl me rs t1 st e
' 18 ths the uir t no 11 t c 1 ee I ed the B s for t RC fl -
cto tr I f on in Secti 3.3 . r4ot 3Ppuch intausS.) l t.sesc.,weri n c yp ._.4m7 ant Catnc.t.mww.3st / 05 04 LG Consistent with industry traveler TSTF-105, the explicit requirements that the RCS flow be measured through the use of a precision heat balance measurement and that the instrumentation used in the performance of the calorimetric flow measurement be calibrated within a specified time period of performing the measurement is moved to a licensee controlled document. The requirement to verify that the RCS flow is within limits remains within the Technical Specification. This is an example of removing unnecessary details from the TS and is acceptable based on the guidance provided in 10 CFR 50.36. 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 l 1s 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. j l
05 06 LS 8 In accordance with NUREG 1431. if any of the DE related i 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 ofChanges to CTS 3M.2 8 5/15/97 i
- - _ - _ _ _ _ - _ _ - _ - _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ - _ = _ _ _ - _ _ _ . . _ _ _ _ - _ - _ _ _ _ _ _ - __ - . _ _ _
INSERT 3A-8 WC 3.2-001 04-11 A A note its 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 L justification is based on the fact that under normal I circumstances, OPTR would not be expected to change significantly within a 12 hour period. If a significant change in OPTR were to occur, it would likely be the result of control rod misalignment ! which would likely be detectable immediately by means of the rod i 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 l requirement to confirm the indication of the remaining three l excore channels. As identified in the NUREG-1431, Rev. I de'inition of QPTR and the SR 3.2.4.2 Bases. OPTR is a ratio of excore detector outputs and that for the purposes of monitoring the OPTR when one power range channel is inoperable, the moveable l incore detectors are used to confirm that the normalized symmetric l power distribution is consistent with the indicated OPTR and any l previous data indicating tilt. l l l L i l 1
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QPTR B 3.2.4 BASES ACTIONS jL1 (continued) If Required Actions A.1 through A.67 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems. l SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 9 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows perfonnance of SR 3.2.4.2 in lieu of SR 3.2.4.1 if wr; tMn ex i;.,,ut fre "w;r ".g 2;tir,n fic C- .;b r.r; ia,,;rer,;;.w%_ confii
. n.m ~+hs . iAAcattim
.,. - 4 he. remamhwe.s.t-This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits.
The Frequency _of 7 days J; . tM C ;h ;; u ^^C"C.: u r.;w,,tsk M;wn ;f tk is ,,rddility tMt thu eh- ; s; Tr. ;.in i.-.,,;rdk N takes 1nto.,ac2uNQ!ggrMcOnMares aya11R1,stoltheioperagisEthtoontmUnbom.
't.s th; """ ;um u i..;,,,;rck, tM Tr;gs., i; ixrand t; 12 b; ;. ";t fr;gx; O :.igt; t; itat my r;kth;ly ;is
;Mr,a in C, kaux Efor those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.
SR 3.2.4.2 This Surveillance is modified by a Note, which states that it 1sinot required ::13 den unt1LKhours~after the input from onem Power Range Neutron Flux channele-erels inoperable and the 1HERMAL POWER is-r > 75% RTP. (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.2 8 3.2-2B 5/15M7
._-____-_________L
Att: chm:nt 3 to WO 98-0078 Pag 31 of 5 l l i l JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS ! CTS 3/4.5 - EMERGENCY CORE COOLING SYSTEMS ITS 3.5 - EMERGENCY CORE COOLING SYSTEMS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES l I l r
Attichm:nt 3 to WO 98-0078 Page 2 of 5 INDEX OF ADDITIONAL INFORMATION I
)
ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER , 3.5.G-1 DC, CP, WC, CA YES l 3.5.1-1 CP NA 3.5.12 WC YES 3.5.1-3 CP NA
- 3.5.1-4 DC NA l 3.5.1-5 DC NA 3.5.1-6 DC, CP, WC, CA YES 3.5.2-1 DC, CP, WC, CA YES l 3.5.2-2 DC, CP, WC, CA YES
'3.5.2-3 DC, CP, WC, CA YES 3.5.2-4 DC, CP, WC, CA YES 3.5.2-5 DC, CP, WC, CA YES 3.5.2 6 DC, CP, WC, CA YES 3.5.2-7 CP NA 3.5.2-8 WC, CA YES 3.5.2-9 DC NA 3.5.3-1 DC, CP, WC, CA YES 3.5.3-2 DC, CP, WC, CA YES .3.5.3-3 DC, CP, WC, CA YES 3.5.3-4 DC, CP, WC, CA YE9 3.5.3-5 DC, CP, WC, CA YES i
3.5.4-1 DC NA 3.5.5-1 DC,CP NA 3.5.5-2 WC, CA YES I CA 3.5-001 DC, WC, CA YES CA 3.5-002 DC, CP, WC, CA YES CA 3.5-003 CA NA CP 3.5-002 - CP NA CP 3.5-003 CP NA CP 3.5-004 CP NA DC ALL-002 (3.5 changes only) DC NA DC 3.5-ED DC NA DC 3.5-001 DC, WC, CA YES DC 3.5-002 DC NA DC 3.5-003 DC NA DC 3.5-005 DC NA j i
Att: chm:nt 3 to WO 98-0078 Pzg3 3 of 5 INDEX OF ADDITIONAL INFORMATION (cont.) ADDITIONAL INFORMATION APPLICABILITY ENCLOSED NUMBER i f DC 3.5-006 DC NA l l TR 3.5-001 DC, CP, WC, CA YES 1 l WC 3.5-ED WC YES l WC 3.5-001 WC YES i WC 3.5-002 WC YES WC 3.5-003 WC YES l i l I I l l l l l 1
Att chm:nt 3 to WO 98-0078 Pege 4 of 5 JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION The following methodology is followed for submitting additional information:
- 1. Each licensee is submitting a separate response for each section.
- 2. If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee), "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.
- 3. If a licensee initiated change does not apply, "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.
- 4. The common portions of the " Additional Information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,3B, 4,6A and 6B of the conversion submittals). The list of attached pages will vary to match j 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. l
- 5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.
- 6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.
- 7. Some changes are not applicable to one licensee but still require changes to the Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one cr 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 additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment request supplement to be provided later.
l l l I i
Attachmsnt 3 to WO 98-0078 Pagn 5 of 5 l JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONALINFORMATION (continued)
- 8. The item numbers are formatted as follows: [ Source][lTS Section)-[nnn) i Source = Q - NRC Question CA- AmerenUE DC-PG&E l WC - WCNOC CP - TU Electric TR - Traveler l
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 l section number. l nnn = a three digit sequential number or ED (ED indicates editorial correction with no impact on meaning) i l l
ADDITIONAL INFORMATION COVER SHEET
. ADDITIONALINFORMATION NO: Q 3.5.G-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 3.5.X Bases General There have been a number of instances ui.J the specific changes to the STS Bases are not properly identified with redline or strikeout marks.
Comment: Perform an audit of all STS Bases markups and identify instances where additions and/or deletions of Bases were not properly identified in the original submittal. FLOG RESPONSE: The submitted ITS Bases markups for Section 3.1 have been compared to the STS Bases. Some differences that were identified were in accordance with the markup methodologies (e.g., deletion of brackets and reviewer's notes). Most of the differences were editorialin nature and would not have affected the review. Examples of editorial changes are:
- 1) Capitalizing a letter with only a " redline" but riot striking out the l lower case letter that it replaced.
- 2) Changing a verb from singular to plural by adding an "s" without
" redlining" the "s".
- 3) Deleting instead of striking-out the A, B, C,.. etc. following a specification title (e.g., SR3.6.6A.7). j
- 4) Changing a bracketed reference (in the reference section) with 1 I
only a " redline" for the new reference but failing to include the strike-out of the old reference.
- 5) In some instances the brackets were retained (and struck-out) but the unchanged text within the brackets was not redlined.
- 6) Not redlining a title of a bracketed section. The methodology calls for the section title to be redlined when an entire section was bracketed.
- 7) Additional text not contained in the STS Bases was added to the l ITS Bases by the lead FLOG member during the development of I the submittal. Once it was determined to not be applicable, the j text was then struck-out and remains in the ITS Bases mark-up.
l Differences of the above editorial nature will not be provided as attachments to this response. The pages requiring changes that are more ! than editorial and are not consistent with the markup methodology are I attached. ATTACHED PAGES: i Encl. 5B B 3.5-28, B 3.5-29 i j/'\ o
l RWST B 3.5.4 BASES I APPLICABLE fraction of the available voltme. The deliverable volme limit is set ' SAFETY ANALYSES by the LOCA and containment analyses. For the RWST, the deliverable (continued) volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minium boron concentration is an explicit assumption in the main l steam line break (MSLB) analysis to ensure the required shutdown capability. W.; i;perte..;e ;f it; vel; i; ; ;11 fer mit; with-e kri,7. inj;;ti= teri ("!T) #.th ; hi#. kra ceri...;r; tin. ForIMIS wit; V.th = OIT ;r r;4 xd O!T krer, r;;ir r.;;. the minima boron concentration limit is an important asstaption in ensuring the required shutdown capability. The maxista boron cuiicantration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures j that the amount of cooling provided from the RWST during the heatup i phase of a line break is consistent with safety analysis ! assumptiori minists is an assumption in both the HSLR and inadvert actuat n analyses, although the inadvert ECCS 1 w.3,sM51 ; hm_ ps actuation o m re]. event is typ'cally nonlimiting. {ceww. rs+pm _ l The MSLB analysis has considered a delay associated with the interlock ' between the VCT and RWST isolation valves, and the results show that e6M WM- the departure from nucle e boiling design basis is met. The delay bracketed "* has been established a 2 seconds, with offsite power available, or 14 15 " 1 M-39 seconds without of site power. This response time includes 2 seconds for electronics delay, a 15 second stroke time fee tc@open the RWST valves, and followediby a 10 second stroke time fee t@close the VCT valves aftergthe.RIESTivalvesareffu11y2open. "l =t; with ; S!T . xd ..;t M sx;rnd with tM d;l;y ;ix; tM "!T Will xpply highly M ;ted .;;t;r pri;r t; '"S :.;it;Lxr. prnid;d tM "!T i; kt xr. th; pap; sd tM sr;. For a large break LOCA analysis, the minimum water voline limit of 2M- ?" .200 39 0 000 gallons and the lower boron concentration limit of 2000 W ppe are used to compute the post LOCA sump boron W.3.5 6as } conce".cration necessary to assure subcriticality. The large break , LOC /. is the limiting case since the safety analysis asstmes that all 1 cuntrol rods are out of the core. (continued) WCGS-Mark-ap ofNUREG-1431 - Bases 3.5 B 3.5 28 5/1SM7
j RWST B 3.5.4 i i l BASES l l l APPLICABLE The upper licit on boron concentration of 2299-2500 ppe is pithin the ) i SAFETY ANALYSES valtas used to determine the maxima allowable time to switch to hot (continued) leg recirculation followina a LOCA._ The_ purpose of switching from cold leg to hot legCn.;ect;= recirculetsc1 Dis to avoid boron precipitation in the core foTTowing the accident.
- newdalAsid k dutWr {0M G-O ,
In tM CCCC .r.:.ly;i;. Inithe minianicoqtaivaant pnessuce-analysis f for EEES3erf_otuunce; evaluation, the containment spray temperature is asstmed to be equal to the RWST lower temperature limit of 95-37'F. If the lower temperature limit is violated, the containment spray
)
further reduces containment pressure, which decreases t k 7.t; et did ;tz ;;r. k .;./d at tM ,rsk coreMoodfrig; rate}and increases peak clad temperature. The upper temperature limit of 100*F l 1s used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break'LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used ; to anximize the total energy release to containment. The RWST satisfies CrMcMon 3 of tM = .";1 icy Ot;t ..t 10 CFR 50;36Tc)I2T(ii). ~~~ v rC'4 qm 2. >.,A. = c4 3.s.co 2. LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment simp to support ECCS and Containment Spray System pump operation in the recirculation mode, j To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs. I APPLICABILITY In MODES 1. 2, 3, and 4. RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since I (continued) l WCGS-Mark-aqp ofNUREG-H31 - Bases 3.5 B 3.5 29 5/15/97
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.1-2 APPLICABILITY: WC REQUEST: CTS 3.5.1 Action b STS 3.5.1 Action B JFD 3.5-2 The CTS markup for 3.5.1 is based on a pending license amendment request (LAR). The change in the completion time for STS Action B is beyond the scope of the conversion review. If the pending LAR is not issued by the time the conversion amendment draft safety evaluation is prepared, the change in the completion time will have to be withdrawn from the conversion submittal. Comment: No action necessary at this time. FLOG RESPONSE: Wolf Creek submitted letter ET 98-0055 on July 17,1998, requesting withdrawal of the portion of the license amendment request that would increase the allowed outage time from 1 hour to 24 hours for an inoperable safety injection accumulator for reasons other than boron concentration riot within limits. However, Wolf Creek expects to resubmit a license amendment request as part of the Westinghouse Owners Group program for relaxation of safety injection accumulator allowed outage times. Wolf Creek is withdrawing the proposed change to the ITS. Should an amendment on the completion time be issued prior to issuance of the conversion amendment draft safety evaluation, a supplement will be issued to the conversion application. ATTACHED PAGES: I' Encl. 2 5-1 ! l Encl.5A 3.5 1 l Encl. 5B B 3.5-6 i Encl.6A 1 Encl.6B 1 ; I i
3/4_5 FMFRGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS
- LIMITING CONDITION FOR OPERATION 3.5.1 hEseh Reactor Coolant System accumulato $shall be OPERABLE.4
. H 'r'
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- d. * . ;n =;_:;:r::: Mt:'-x .???:rd22:;f; APPLICABILITY: MODES 1,2, and 3*.
ACTION
- a. With{three accumulators OPERABLE andone accumulator inoperable, Mff!46AUP due to boron concentration not within limits, either restore the boron concentration EM'N to within theebewe limits within 72 hours or be in at least HOT STANDBY within the next 6 hours, and reduce RCS pressure to less than[or equal tol1000 psig {}f{ff"ED.j$5g~
within the following 6 houts.
- b. With(three accumulators OPERABLE an3one accumulator inoperable for [$1MA?9 reasons other_than ACTION a., restore the inoperable accumulator to OPERABLE "h**
status withirG!PpprSor be in at least HOT STANDBY within the next 6 hours and reducerRCs pressure to less tharfor equal to]1000 psig within the following I'" ED' E"E"
'**"l""
6 hours i IO Q 3.5.1-2.,1 SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE: l l
- a. At least once per 12 hours by:
- 1) Verifying that the contained borated water volume [is 2 6122 gallons) fand s 6594 gallons)and nitrogen cover-pressure in the tanks
- . ~^^ ?.::: ' - .Q 2 585 poig and s 666 ps6g]and
- 2) Venfying that each accumulator isolation valve is open.
g r--;ter pressure above 1000 psig linl4MX3I3 anim.iwamA WOLF CREEK- UNIT 1 3/4 6-1 Amendment No. 44,24103 Markup ofCTS3N.5 5/15/97 i
Acetmulators 3.5.1 3.5 EMERGENCY C0RE COOLING SYSTEMS (ECCS)
~
3.5.1 Accumulators LCO 3.5.1 Feur ECCS accumulators shall be OPERABLE. [ggggy APPLICABILITY: MODES I and 2, MODE 3 with pr;;;;;;ri;;r RCS pressure > 1800 psig. ff$$$$d ACTIONS COWITION REQUIRED ACTION COMPLETION TIE A. One accumulator A.1 Restore boron 72 hours inoperable due to boron concentration to within concentration not within limits, limits. p s.s. i-2. , B. One accumulator inoperable for reasons B.1 Restore accumulator to OPERABLE status. hour @ % other than Condition A. C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B &lil . _ i not met. $$$.55U1!$ C.2 Reduce pr;;;;;;ri;;r RCS 12 hours pressure to $$$$8M_s. s 1000 psig. 4 D. Two or more accumulators D.1 Enter LCO 3.0.3. Innediately inoperable. I ( WCGS-Mark-up ofNUREG-H31-1753.5 3.5 1 5/158 7
Acetmulators B 3.5.1
. BASES ACTIONS U (continued) precipitation time may be reduced. The boron in the accumulator contributes to the asstaption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of, the core subcritical.
One accumulator below the minima boron concentration limit, however, ; will have no effect on available ECCS water and an insignificant ; effect on core subcriticality during reflood. Boiling of ECCS water j in the core during reflood concentrates boron in the saturated liquid that remains in the core. In c.ddition. ;;.rr;nt m; lysis t;J .iqua d;.~a;tret; tret tre aca=1;ter; de ret di;;terg felle.; ins ; lers; min st;n lire bruk for the mjerity ;f pl=ts. Den if the3 de dixter9; If"theinecumulatorsidischarge follow 1.nga0arge;, main! steam 11nelbtwakytth;offsite power'available, their impact is minor and not a design limiting event. Thus. 72 hours is allowed to return the boron concentration to within limits. IL1 If one accumulator is inoperable for a reason other than boron gh concentration within*1 heur[, ___r the acc'=>1ator_ must be returnedato to OPERABL
.e an, .1 feff Z Thi
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- _ _Wi.. p'Tito f ~ t' _ ; s? ified ing- unavaildflit" ofiUO8icirs" ~ 3.5.l-1
. n this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severit of the consequences should a LOCA occur in these conditions, t ur Completion Time to open the valve, remove power to the 3,,,,.q valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions.
C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant (continued) WCGS-Mark-up ofNUREG-1431 - Bases 3.5 B 3.5 6 S/158 7
i DIFFERENCES FROM NUREG 1431
.Section 3.5 i 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 merk ups. For enclosures 3A, 38. 4, 6A and 6B, text in brackets "[ ]" indicates the information is ' plant specific and is not common to all the Joint Licensing Subconnittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUMBER JUSTIFICATION 3.5 1 Replaces reference to the " pressurizer pressure" with a reference to the "RCS pressure" in the APPLICABILITY, Required Action C.2, and SR 3.5.1.5. Required Action C.2 requires reducing pressurizer pressure to less than 1000 psig. However, pressurizer pressure instrumentation does not have the range to read that pressure. Consequently RCS pressure instrumentation is used. For the purposes of this LCC, the use of RCS pressure is equivalent. This is consistent with traveler TSTF-117. g l 3.5 2 fThe lee T of 3. . , TI is c ngedd 1 l ' ho to ho to fl he c ent as [ spdby l cen a t eques . Compl ion Ti ch 1 i s rt ya valu on o core ge fr ue base,on d jccumu or vail 111ty The r sults - n ip!Iignif cant
'iner se CDF f all out time well n egess o 24 .
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3.5 3 Adds the word " mechanical" with regard to throttle valve position stop consistent with the current TS. These valves have l mechanical stops that maintain the valves in position for proper ECCS performance. 3.5-4 Not applicable to WCGS. See Conversion Comparison Table ! (Enclosure 6B). 1 3.5-5 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 68). i 3.5 6 SR 3.5.3.1 Note (An RHR train may be considered OPERABLE during alignment and operation for decay heat removal, if capable of l j being realigned to the ECCS mode of operation) is moved to the ' LCO per traveler TSTF-90. The Note is more appropriately placed , into the LCO because it defines the intended capability of the ECCS equipment. l l WCGS-Differencesfrom NUREG-1431 - ITS 3M.S 1 S/158 7
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ADDITIONAL INFORMATION COVER SHEET I ADDITIONAL INFORMATION NO: O 3.5.1-6 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 1-07 LG CTS 4.5.1.1.b (DC, CA, WC) CTS 4.5.1.b (CP) ITS SR 3.5.1.4 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG RESPONSE: DOC 1-07 LG has been revised to provide additional justification for the proposed change by adding the following information:
"The RWST has its own LCO and SRs to verify OPERABILITY and cross-references to other specifications are generally inconsistent with the ITS format and are not required to impose OPERABILITY on the referenced equipment. The RWST boron concentration is maintained between
[2350] ppm and [2500] ppm, which is higher than the minimum boron concentration required to be maintained in the accumulators. If there were reason to doubt the RWST boron concentration, ITS 3.5.4 Condition A would be entered with its 8 hour Completion Time. In addition, ITS SR 3.5.4.3 verifies the boron concentration of the RWST every 7 days. Therefore, it is unlikely that the boron concentration being added to the accumulators would be below [2350] ppm. Additionally, plant procedures implementing the SR 3.5.1.4 Bases specify that if the RWST has been diluted since its last boron concentration sample per SR 3.5.4.3, the boron concentration in the accumulators must be verified within 6 hours after adding [70) gallons or more to the accumulators from the RWST. The moving of this detail to the ITS Bases maintains consistency with NUREG-1431 and is not necessary to adequately protect the health and safety of the public. Details for performing surveillance requirements are more appropriately specified in the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the plant conditions appropriate to perform a surveillance test is an issue for procedures and scheduling and has been previous!y determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is cons! stent with the vast majority of other surveillance requirements that do not dictate plant conditions for surveillance. Any change to this detail will be made in accordance with the Bases Control Program described in ITS ; Section 5.5.14." ! ATTACHED PAGES: 1
- Encl. 3A 1, 2 i
I
DESCRIPTION OF CHANGES TO CURRENT TS SECTION 3/4.5
, This enclosure contains a brief description / justification for each marked up change to the current Technical Specifications. The changes are identified by change numbers contained in enclosure 2 (Mark up of the current Technical Specifications).
In addition, the referenced No Significant Hazards Considerations (NSHCs) are contained in enclosure 4. Only technical changes are discussed: administrative changes (i.e., format, presentation, and editorial changes) made to conform to NUREG 1431 Revision 1 are not discussed. For enclosures 3A, 3B, 4, 6A and 6B, text in brackets "[ ]" indicates the information is plant specific and is not common to all the Joint Licensing Subcommittee (JLS) plants. Empty brackets indicate that other JLS plants may have plant specific information in that location. CHANGE NUMBER H2iG DESCRIPTION ! 1 01 M Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 1 02 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 1 03 A Replaces reference to the " pressurizer pressure" with a reference to the "RCS pressure". ACTIONS a and b require reducing pressurizer pressure to less than 1000 psig. However, pressurizer pressure instrumentation does not have the range to read that pressure. Consequently RCS pressure instrumentation is used. For the purnoses of this LCO, the use of RCS pressure is equivalent. 1-04 LS 8 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 1 05 LS 9 Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 1 06 A Adds the words "with 3 accumulators OPERABLE and" to both Action statements to make entry into LC0 3.0.3 mandatory with two or more accumulators inoperable. This change is consistent with NUREG-1431. Rev.1, and is considered administrative in nature since it reflects current plant practice, i.e., current ACTION Statements a and b are not entered at the same time on different accumulators. 1-07 LG The SR currently requires a 6 hour surveillance if the makeup source is the RWST and the RWST has been diluted since verifying its boron concentration per the RWST LCO. The proposed change would move the statement "and the RWST has not been diluted since_verifyint... to the__ITS SR 3.5.1.4 Bases. QEi: lac 1 of pt i1 5 =t WEM p o3,s.i.4l WCGS-Description of Changes to CTS 3M.S 1 S/1S/97
(. l CHANGE NUl6ER !GjC DESCRIPTION
$ o z.s. -g ]
1 08 A Not applicable to WCGS. See Conversion Comparison Table ] (Enclosure 3B). 2 01 LG Consistent with NUREG 1431 Rev 1, the LCO and ACTION a are revised to replace the word " subsystem" with the word
" train" and the descriptive information in the LCO is ,
moved to the BASES. Whereas, there is no technical change l associated with t replacement of the term " subsystem," j
" train" better des ribes that all parts of the required l system (e.g., pipi g, instruments, controls, etc.) must be operable to suppo t th reguired safety functions.
Sissa.y '3A . 2b)--t a s .s .z . J ; 2 02 LS 1 Consistent with NUREG 1431 Rev 1, a note with respect to RCS pressure isolation valve testing is added to the LCO. Plant design requires closure of certain valves in SI injection paths to perform PIV testing. Isolation of the ) injection paths in H0DE 3 is currently prohibited as it would constitute entering TS 3.0.3 since both SI trains ! would be made administratively inoperable. In actuality, the flow paths are readily restorable from the control ; room, and a spurious single active failure is not likely in the short term (2 hours). The new note will allow closing these valves without declaring either SI train inoperable. This change is consistent with traveler TSTF-153. 2 03 LS 2 This change revises Action a to allow for increased flexibility in plant operations under circumstances where components in opposite trains are inoperable, but at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available. Due to the design of the ECCS subsystems, the inoperable condition of one or more components in each train does not necessarily render the ECCS inoperable for performing its safety function. The allowed outage time of 72 hours is unchanged; but, it is to be contingent on being capable of providing 100% of the ECCS flow equivalent to a single operagble ECCS train. This change is consistent with NUREG-1431. Rev.1. 2 04 TR 2 Consistent with NUREG 1431 Rev.1, the requirement to submit a Special Report within 90 days of an ECCS actuation and injection event is deleted. This change is WCGS-Description of Changes to CTS 3M.5 2 S/158 7
l INSERT 3A-2a 0 3.5.1-6 , i The RWST has its own LCO ar.d SRs to verify OPERABILITY and cross-references to other specifications are generally inconsistent with the ITS format and are not required to impose OPERABILITY on the referenced equipment. The RWST boron concentration is maintained between [2350] ppm and [2500] ppm which is higher than the minimum boron concentration required to be maintained in the accumulators. if there were reason to doubt the RWST boron concentration, ITS 3.5.4 Condition A would be entered with its 8 hour Completion Time. In addition. ITS SR 3.5.4.3 verifies the boron concentration of the RWST every 7 days. Therefore, it is unlikely that the boron concentration being added to-the accumulators would bq below [2350] ppm. Additionally, plant procedures implementing the SR 3.5.1.4 Bases specify that if 'the RWST has been diluted since its last baron concentration sample per SR 3.5.4.3, the boron concentration in the accumulators must be verified within 6 hours after adding [70] gallons or more to the accumulators from'the RWST. The moving of thi.s detail to the ITS Bases maintains consistency with NUREG-1431 and_is not necessary to adequately protect the health and safety of the public. Details for performing surveillance requirements' are more appropriately specified in the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the plant conditions appropriate to perform a surveillance test'is an issue for procedures and scheduling and has been_ previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is consistent-with the vast majority of other j surveillance requirements that do not dictate plant conditions for surveillance. Any change to this detail will be made in accordance with the Bases Control Program described in ITS Section 5.5.14. INSERT 3A-2b Q 3.5.2-1 l The proposed change is consistent rith NUMARC 93-03. " Writer's Guide for l the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the LCO describes as simply as possible the lowest I functional capability of the system and relegates the details of what constitutes an OPERABLE systei to the Bases. Therefore, the details of what constitutes an OPERABLt subsystem (train) such as required pumps. heat exchangers and flow paths, are more appropriately discussed in the l bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.2-1 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 2-01 LG CTS 3.5.2 LCO ITS 3.5.2 LCO The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail is not necessary in the ITS. FLOG RESPONSE: DOC 2-01-LG has been revised to provide additionaljustification for the proposed change by adding the following information:
"The proposed change is consistent with NUMARC 93-03, " Writer's Guide for the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the LCO describes as simply as possible the lowest functional capability of the system and relegates the details of what constituted an OPERABLE system to the Bases. Therefore, the details of what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the Bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY."
ATTACHED PAGES: Encl. 3A 2 l
q
- - - - - - - - - ~ - - - - - - - - - - - - -- - - - --- - -- --- -- - - - - - -- -
r--------- Conse IDGER NE DESCRIPTION
-/ _
l the I anyt!f co st wi he nd nforelitsen ! aine rin t a . IMS*LT SA - ZA -
-_- v Q 2.5.1 -G ]
1 08 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38), 2 01 LG Consistent with NUREG 1431 Rev 1. the LCO and ACTION a are
- revised to replace the word " subsystem". with the word
! " train" and the descriptive information in the LCO is i moved to the BASES. Whereas, there is no technical change-associated with t replacement of the term " subsystem,"
" train" better des ribes that all parts of the required l system (e.g., pipi ,-instruments, controls, etc.) must be operable to suppo t t r uired safety functions.
W r M- 2.ig)-.- L o 3.s . 2. - 1 J 2 02 LS 1 Consistent with NUREG 1431 Rev 1 a note with respect to RCS pressure isolation valve testing is added to the LCO. Plant design requires closure of certain valves in SI ! injection paths to perform PIV testing. Isolation of the l injection paths in MODE 3 is currently prohibited as it would constitute entering TS 3.0.3 since both SI trains i would be made administratively inoperable. In actuality, i the flow paths are readily restorable from the control room, and a spurious single active failure is not likely in the short term (2 hours). The new note will allow closing these valves without declaring either SI train inoperable. This change is consistent with traveler TSTF-153. 2 03 LS 2 This change revises Action a to allow for increased flexibility in plant operations under circumstances where components in opposite trains are inoperable, but at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available. Due to the design of the ECCS subsystems, the inoperable condition of one or more l components in each train does not necessarily render the ECCS inoperable for performing its safety function. The allowed outage time of 72 hours is unchanged: but, it is to be contingent on being capable of providing 100% of the ECCS flow equivalent to a single operagble ECCS train. This change is consistent with NUREG-1431, Rev. 1. L 2 04 TR 2 Consistent with NUREG 1431 Rev. 1, the requirement to submit a Special Report within 90 days of an ECCS actuation and injection event is deleted. This change is WCGS-Description ofChanges to CTS 3M.S 2 SMSM7 i
INSERT 3A-2a 0 3.5.1-6 l The RWST has its own LCO and SRs to verify 0PERABILITY and cross-references to other specifications are generally inconsistent with the ITS format and are not required to impose OPERABILITY on the referenced equipment. The RWST boron concentration is maintained between [2350] ppm and [2500] ppm, which is higher than the minimum boron concentration required to be maintained in the accumulators. If there were reason to doubt the RWST boron concentration, ITS 3.5.4 Condition A would be entered with its 8 hour Completion Time. In addition, ITS SR 3.5.4.3 verifies the boron concentration of the RWST every 7 days. Therefore, it is unlikely that the boron concentration being added to the accumulators would be below [2350] ppm. Additionally, plant procedures implementing the SR 3.5.1.4 Bases specify that if the RWST has been diluted since its last boron concentration sample per SR 3.5.4.3, the boron concentration in the accumulators must be verified within 6 hours after adding [70] gallons or more to the accumulators from the RWST. The moving of this detail to the ITS Bases maintains consistency with NUREG-1431 and is not necessary to adequately protect the health and safety of the public. Details for performing surveillance requirements are more appropriately specified in the plant procedures required by ITS l 5.4.1 and the ITS Bases. Control of the plant conditions appropriate to perform a surveillance test is an issue for procedures and scheduling l and has been previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this , l licensee control is consistent with the vast majority of other j surveillance requirements that do not dictate plant conditions for i surveillance. Any change to this detail will be made in accordance with the Bases Control Program described in ITS Section 5.5.14. INSERT 3A-2b 0 3.5.2-1 i The proposed change is consistent with NUMARC 93-03, " Writer's Guide for ' the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the LC0 describes as simply as possible the lowest functional capability of the system and relegates the details of what i constitutes an.0PERABLE system to the Bases. Therefore, the details of L what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore. the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY.
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.2-2 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 2-09 LG CTS 4.5.2.c The referenced DOC describes the change to the CTS but does not provide any justification for making the change othe, than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this surveillance is not necessary in the ITS. FLOG RESPONSE: DOC 2-09 LG has been revised to provide additionaljustification for the, proposed change by adding the following information:
" CTS SR 4.5.2.c requires a visual inspection to verify that no loose debris is present in the containment which could be transported to the containment sump and cause restriction to the pump uuction during LOCA conditions at the frequency specified. This ensures tilat during the process of performing maintenance or other work inside containment that debris is appropriately discarded. Existing procedures restrict containment entries and assure accountability of items entering containment such that they are removed at the completion of the containment entry. ITS SR 3.5.2.8 continues to require a visualinspection every 18 months on each of the ECCS train containment sump suction inlets to ensure that the sump suction inlet is not restricted by debris.
Therefore, this detail is not required to be in the technical specifications and moving this requirement maintains consistency with NUREG-1431." ATTACHED PAGES: Encl. 3A 3 i I l ) f L
i CHANGE
!MBfE li2lC DESCRIPTION acceptable because the requirement to submit a report is sufficiently addressed by the reporting requirements contained in 10CFR50.73.
2 05 LS 3 This change revises the LCO applicability note to allow operation in MODE 3 pursuant to LC0 3.5.3 until "all" cold legs exceed the RCS temperature setpoint in lieu of "one or more." The previous allowance was [within 4 hours or prior to the temperature of "one or more" of the RCS cold legs exceeding 375T, whichever comes first). The 375Y is a nominal _ temperature selected to give time to restore the pump operability without delaying startup. The four hour limit is unchanged. Changing "one or more" to "all" is still bounded by the 4 hour limit. This change is consistent with NUREG 1431 Rev. 1. 2 06 Not used.
- 2 07 A Consistent with NUREG 1431 Rev.1. this change revises the surveillance to make it clear that " listed" valve position L
is the concern and not indicated position in the control
- room. The surveillance can be satisfied using indicated position in the control room but may also be satisfied using local observation. This is an administrative change since the surveillance ~ acceptance criteria are not L changed.
, 2 08 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 2 09 LG The visual inspection surveillance performed when establishing containment integrity is moved __ to a licensee _ co.Lntrolled document./Mov' th r orroqGira.m is] [seffsisjefit wgNgG- -- 1 lesq SA-L /
%Q 3.5.2-2.]
~ -
2 10 A Consistent with NUREG 1431 Rev.1. the current TS SR for verifying interlock action of the RE system is moved to improved TS SR 3.4.14.2. ' 2 11 TR 1 Consistent with NUREG 1431 Rev. 1. the ECCS pump and valve actuation SR is changed to allow the use of an actual signal, if and when one occurs, to satisfy surveillance requirements. The specific signals used to actuate the pumps and valves have been moved to the Bases. ' O h T 3A _3 @ l Q .s.s. 2. - 31 2 12 LG The ECCS pump performance is revised to be consistent with NUREG 1431 Rev.1. The test method and specific data required to verify pump performance are moved to licensee controlled documents. Specification 4.0.5 no longer exists WCGS-Description of Changes to CTS 3M.S .3 SMSM7
INSERT 3A 3a 0 3.5.2-2 CTS SR 4.5.2.c requires a lisu.al inspection to verify that no loose debris is present in the containment which could be transported to the containment sump and cause restriction to the pump suction during LOCA i I conditions at the frequency specified. This ensures that during the process of performing maintenance or other work inside containment that debris is appropriately discarded. Loose debris in the containment does not directly render the ECCS inoperable or incapable of performing its intended function. Existing procedures restrict containment entries and assure accountability of items entering containment such that they are i removed at the completion of the containment entry. ITS SR 3.5.2.8 i continues to require a visual inspection every 18 months on each of the ECCS train containment sump suction inlets to ensure that the sump suction inlet is not restricted by debris. Therefore this detail is not required to be in the technical specifications and moving this j requirement maintains consistency with NUREG-1431. l l INSERT 3A-3b 0 3.5.2-3 In several specifications throughout the CTS. OPERABILITY of certain equipment is demonstrated by ensuring that the equipment performs its safety function upon receipt of a simulated test signal. The intent of l a ' simulated
- signal was to be able to perform the required testing without the occurrence (or without causing) an actual signal generating event, However, the unintended effect was to require the performance of the surveillance (using a test signal) even if an actual signal had previously verified the operation of the equipment. This change allows credit to be taken for actual events when the required equipment actuates successfully.
l While the occurrence of events that cause actuation of accident mitigation equipment is undesirable. the actuation of mitigation equipment on an actual signal is a better demonstration of its OPERABILITY than an actuation using a test signal. Thus, the change does not reduce the reliability of the equipment tested. The change also improves plant safety by reducing the amount of time the equipment is taken out of service for testing. and thereby increasing its availability during an actual event. and by reducing the wear of the equipment caused by unnecessary testing. i
ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.5.2-3 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 2-11 TR-1 CTS 4.5.2.e ITS SR 3.5.2.5 & SR 3.5.2.6 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: The NSHC for this change appears to provide the needed justification. Therefore, please incorporate the information contained in the NSHC into the subject DOC. FLOG RESPONSE: The CTS requires the use of a test signal for initiation of valid tests. The unintentional result was to require the performance of the verification even if an actual signal has already verified proper operation of equipment. TR-1 allows either an actual or test signal. DOC 2-11 TR-1 has been revised to provide additional discussion to allow the use of an actual signal to meet this surveillance requirement. ATTACHED PAGES: Encl. 3A 3 l-l
l CHANGE NUPBER NStE DESCRIPTION l acceptable because the requirement to submit a report is sufficiently addressed by the reporting requirements contained in 10CFR50.73. 1 2 05 LS 3 This change revises the LCO applicability note to allow operation in MODE 3 pursuant to LCO 3.5.3 until "all" cold legs exceed the RCS temperature setpoint in lieu of "r e or more " The previous allowance was [within 4 hours or prior to the temperature of "one or more" of the RCS cold legs exceeding 375T. whichever comes first). The 375T is a nominal temperature selected to give time to restore the pump operability without delaying startup. The four hour limit is unchanged. Changing "one or more" to "all" is still bounded by the 4 hour limit. This change is consistent with NUREG 1431 Rev. 1. 2 06 Not used. l 2 07 A Consistent with NUREG 1431 Rev.1. this change revises the surveillance to make it clear that " listed" valve position is the concern and not indicated position in the control room. The surveillance can be satisfied using indicated i position in the control room but may also be satisfied using local observation. This is an administrative change . since the surveillance acceptance criteria are not changed. 2-08 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 2 09 LG The visual inspection surveillance performed when establishing containment inte<4rity is moved to _a licensee i.rtrolled document.1 t orregridi. . 6. is
@ sis p g % G 1 NSqC 3A-3A
%3.s.2 21 2 10 A Consistent with NUREG-1431 Rev.1. the current TS SR for verifying interlock action of the RHR system is moved to improved TS SR 3.4.14.2.
2 11 TR 1 Consistent with NUREG 1431 Rev.1. the ECCS puso and valve actuation SR is changed to allow the use of an actual signal if and when one occurs, to satisfy surveillance requirements. The specific signals used to actuate the i pumns and valves have been moved to the Bases. ' Oi5 EAT @C-SkD
~
1 o.s.s.2.-3 1 2 12 LG The ECCS pump performance is revised to be consistent with NUREG 1431 Rev.1. The test method and specific data required to verify ptmp performance are moved to licensee controlled documents. Specification 4.0.5 no longer exists WCGS-Description of Changes to CTS 3N.S 3 S/2S/97 I 1
INSERT 3A-3a 0 3.5.2-2
. CTS SR 4.5.2.c requires a visual inspection to verify that no loose debris is present in the containment which could be transported to the containment sump and cause restriction to the pump suction during LOCA conditions at the frequency specified. This ensures that during the process of performing maintenance or other work inside containment that debris is appropriately discarded. Loose debris in the containment does not directly render the ECCS inoperable or incapable of performing its intended function. Existing procedures restrict containment entries and assure accountability of items entering containment sucn that they are removed at the completion of the containment entry. ITS SR 3.5.2.8 continues to require a visual inspection every 18 months on each of the ECCS train containment sump suction inlets to ensure that the sump suction inlet is not restricted by debris. Therefore, this detail is not required to be in the technical specifications and moving this requirement maintains consistency with NUREG-1431.
INSERT 3A-3b 0 3.5.2-3 In several specifications throughout the CTS. OPERABILITY of certain equipment is demonstrated by ensuring that the equipment performs its safety function upon receipt of a simulated test signal. The intent of a ' simulated
- signal was to be able to perform the required testing without the occurrence (or without causing) an actual signal generating event. However, the unintended effect was to require the performance of the surveillance (using a test signal) even if an actual signal had previously verified the operation of the equipment. This change allows credit to be taken for actual events when the required equipment actuates successfully.
While the occurrence of events that cause actuation of accident mitigation equipment is undesirable, the actuation of mitigation equipment on an actual signal is a better demonstration of its OPERABILITY than an actuation using a test signal. Thus, the change does not reduce the reliability of the equipment tested. The change also improves plant safety by reducing the amount of time the equipment is taken out of service for testing, and thereby increasing its availability during an actual event, and by reducing the wear of the equipment caused by unnecessary testing, i 1
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.2-4 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 2-12 LG CTS 4.5.2.f ITS SR 3.5.2.4 The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS.
~
Comment: Please revice the DOC to include additionaljustification as to why this detail is not necessary in th; ITS. FLOG RESPONSE: DOC 2-12-LG has been revised to provide additional justification for the proposed change by adding the following information:
"lTS SR 3.5.2.4 retains the SR requirement and references the Inservice Testing Program (IST), discussed in ITS 5.5.8, for the surveillance Frequency. The specific SR acceptance criteria for the pumps have been l moved to the ITS SR 3.5.2.4 Bases. Although this may make the ECCS I pump performance testing more flexible in the future, only with regard to licensee control over the numerical values of the acceptance criteria, this testing must continue to conform to the IST program requirements.
Revisions to the acceptance criteria will have to meet the requirements of the Bases Control Program discussed in ITS 5.5.14. Details for performing surveillance requirements are more appropriately specified in i the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the acceptance criteria for a surveillance test is an issue for the IST procedures and has been previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is consistent with the vast majority of other surveillance requirements that do not dictate plant conditions for surveillance." ATTACHED PAGES: Encl.3A 4 l 1
CHANGE NUMBER E2iG DESCRIPTION in the improved TS: however, the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 of the improved TS. The IST Program is referenced directly for the frequency of testing. hT 3A} 4@ o.s.s.z-4] 2 13 TR 3 The CTS allowance, which permits the ECCS throttle valves to be declared OPERABLE without verifying ECCS throttle valve stop position for 4 hours following valve stroke testing or maintenance is deleted from the current TS consistent with NUREG 1431 Rev 1. The ECCS throttle valves are manual valves and plant procedures governing post-maintenance test requirements specify verification of correct throttle position prior to declaring the valves OPERABLE. Explicit post-maintenance TS surveillance requirements have been deleted because these requirements are adequately addressed by administrative post-maintenance programs. 2 14 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 2 15 LG The surveillance requirement for the flow balance test following ECCS modifications is moved to a licensee controlled document. This requirement is not included in NUREG 1431 Rev. 1. [gf 34 4b Q 3.s.2.-s l 2-16 LG Tha specific means by which the ECCS piping is assured to be full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with the kind of information contained in the Bases.
@ r'-5T:R-)--l o s.s.2 - s.1 2 17 A Adds the phrase "that is not locked sealed or otherwise secured in position" with regard to which valves require actuation testing. This change is merely a clarification.
Valves that are secured in place are secured in the position required to meet their safety function. The actuation testing ensures that valves can move to the position that meets their safety function. If the valves are secured in the position that meets their safety function no testing is necessary. 2 18 LG Not applicable to WCGS. See Conversion Comparison Table , (Enclosure 3B). . I 2-19 LG Consistent with NUREG 1431 Rev. 1 this change moves the requirement that the 18 month verification of automatic ECCS valve actuation and ECCS pump actu_ation be performed during shutdown to the Bases. QERT JAj4a].Q3.5.2-el WCGS-Description of Changes to CTS 3M.S 4 5/15/97
INSERT 3A 4d 0 3.5.2-4
~
ITS SR 3.5.2.4 retains the SR requirement and references the Inservice Testing Program (IST), discussed in ITS 5.5.8, for the surveillance Frequency. The specific SR acceptance criteria for the pumps have been moved to the ITS SR 3.5.2.4 Bases. Although this may make the ECCS pump performance testing more flexible in the future, only with regard to licensee control over the numerical values of the acceptance criteria. this testing must continue to conform to the IST program requirements. Revisions to the acceptance criteria will have to meet the requirements of the Bases Control Program discussed in ITS 5.5.14 Details for performing surveillance requirements are more appropriately specified in the plant procedures required by ITS 5.4.1 and the ITS Bases. Control of the acceptance criteria for a surveillance test is an issue for the IST procedures and has been previously determined by NRC to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this licensee control is consistent with the vast majority of other surveillance requirements that do not dictate plant conditions for surveillance. i I
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.2-5 APPLICABILITY: DC, CP, WC, CA REQUEST: DOC 2-15 LG CTS 4.5.2.h The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this surveillance is not necessary in the ITS. FLOG RESPONSE: DOC 2-15 LG has been revised to include additionaljustification as to why this surveillance is not necessary in the ITS.
]
ATTACHED PAGES: Encl. 3A 4 1 l 1 2 l 9
I CHANGE NUMBER 82iG DESCRIPTION in the improved TS: however. the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 1 of the improved TS. The IST Program is referenced directly for the frequency of testing. Q t 34}. 4 M o.s.s.2.-4) 2 13 TR 3 The CTS allowance, which permits the ECCS throttle valves to be declared OPERABLE without verifying ECCS throttle valve stop position for 4 hours following valve stroke testing or maintenance is deleted from the current TS consistent with NUREG 1431 Rev 1. The ECCS throttle valves are manual valves and plant procedures governing post-maintenance test requirements specify verification of correct throttle position prior to declaring the valves OPERABLE. Explicit post maintenance TS surveillance requirements have been deleted because these requirements are adequately addressed by administrative post-maintenance programs. 2-14 A Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 2-15 LG The surveillance requirement for the flow balance test i following ECCS modifications is moved to a licensee ! controlled document. This requirement is not included in NUREG 1431 Rev. 1. gf 34 -4M Q 3.s.2.-s l 2 16 LG The specific means by which the ECCS piping is assured to be full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with the kind of information contained in the Bases. Wer 34 Q-1 o s.s.2.-5 1 2 17 A Adds the phrase "that is not locked, sealed, or otherwise secured in position" with regard to which valves require actuation testing. This change is merely a clarification. Valves that are secured in place are secured in the position required to meet their safety function. The actuation testing ensures that valves can move to the position that meets their safety function. If the valves are secured in the position that meets their safety function, no testing is necessary. 2 18 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). I 2 19 LG Consistent with NUREG 1431 Rev. 1 this change moves the I requirement that the 18 month verification of automatic ECCS valve actuation and ECCS pump actuation be performed during shutdown to the Bases. $ SEW.T 5AMQ3.5.2-el WCGS-Description of Changes to CTS 3M.5 4 SAS/97
INSERT 3A-4a 0 3.5.2-8 Such limitations are not required to be detailed in the Technical Specifications. These surveillance requirements are typically performed
- during plant shutdown, however, if for instance, an actual signal is
! generated while operating, results should be useable even though the plant is not " shutdown." Similarly, if testing would be required to l complete some repair or modification made while operating, a shutdown should not be required. Therefore, the CTS wording is consistent with the NUREG-1431 Bases wording that states: The 18 month Frequency is based on the need to perform these Surveillance under the conditions
- that apply during a plant outage and the potential for an unplanned l transient if the Surveillance were performed with the reactor at power.
Moving these details to the Bases is consistent with other requirements specified in NUREG-1431, Rev. 1. This change moves to the Bases details that are not necessary to provide operational safety wnile retaining in technical specifications the basic requirements for maintaining OPERABILITY. INSERT 3A-4b 0 3.5.2-5 l Plant procedures governing the restoration of equipment after l maintenance specify the requirements for determining the appropriate post maintenance testing. Any time the Operability of a system or L component has been affected by repair, maintenance, or replacement of a l component, post maintenance testing is required to demonstrate ( .0perability of the system or component. As such, the requirement to l perform a flow balance test after modifications that alter ECCS l subsystem flow' characteristics is not required to be in the TS to l provide adequate protection of the public health and safety. This requirement has been moved to the [U]SAR (for Callaway, Diablo Canyon, and Wolf Creek) or TRM (Commanche Peak). These licensee. controlled documents containing the moved requirements will be maintained using the provisions of 10 CFR 50.59. Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner, L 1 INSERT 3A-4c 0 3.5.2-6 l The requirements of ITS LC0 3.5.2 and the associated Surveillance l Requirements are adeauate to ensure the ECCS are maintained OPERABLE. l As a result, the methods of performing Surveillance are not necessary to ensure the ECCS can perform their intended rafety function and the details are not required to be in the TS to provide adequate protection of the public health and safety. The 1TS Base!. containing the moved requirements will be maintained using the provisions of 10 CFR 50.59, as required by the Bases Control Program described in ITS Section 5.5.14. Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner. l
ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.2-6 APPLICABILITY: DC, CP, WC, CA l REQUEST: DOC 2-16 LG CTS 4.5.2.i ITS SR 3.5.2.3 The referenced DOC describes the change to the CTS but does not provide any , justification for making the change other than that it is consistent with the STS. l Comment: Please revise the DOC to include additionaljustification as to why this detail J is not necessary in the ITS. FLOG RESPONSE: DOC 2-16 LG has been revised to include additional justification as to why this detail is not necessary in the ITS. ATTACHED PAGES: Encl. 3A 4 l f o___-------____-------- -__- - -
(_ CHANGE NUMBER tGlC DESCRIPTION
, in the improved TS: however, the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 of the improved TS. The IST Program is referenced directly for the frequency of testing.
hT 3A[4M QJ.E.2-4l 2 13 TR 3 The CTS allowance, which permits the ECCS throttle valves to be declared 0FERABLE without verifying ECCS throttle < valve stop position for 4 hours following valve stroke I l testing or maintenance is deleted from the current TS f consistent with NUREG 1431 Rev 1. The ECCS throttle valves are manual valves and plant procedures governing post-maintenance test requirements specify verification of correct throttle position prior to declaring the valves OPERABLE. Explicit post maintenance TS surveillance requirements have been deleted because these requirements are adequately addressed by administrative post-maintenance programs. 2 14 A Not applicable to WCGS. See Conversion Comparison Table I (Enclosure 38). '~ 2 15 LG The surveillance requirement for the flow balance test following ECCS modifications is moved to a licensee l controlled document. This requirement is not included in 1 NUREG 1431 Rev. 1. @segf 34 4g q 3,g,2_,[ ) 2 16 LG The specific means by which the_ ECCS piping is assured to l be full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with ; the kind of information contained in the Bases. Faf sa =D--Lo S 5 2 1 2 17 A Adds the phrase "thit is not locked, sealed, or otherwise secured in position" with regard to which valves require actuation testing. This change is merely a clarification. Valves that are secured in place are secured in the position required to meet their safety function. The actuation testing ensures that valves can move to the j position that meets their safety function. If the valves , are secured in the position that meets their safety function, no testing is necessary. 2 18 LG Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 3B). 1 2 19 LG Consistent with NUREG 1431 Rev. 1. this change moves the requirement that the 18 month verification of automatic ECCS valve actuation and ECCS pump actuation be performed during shutdcwn to the Bases, fT]JA {4@Q 3.5.2-el WCGS-Description ofChanges to CTS 3M.S 4 5/1587 l L_-_____
INSERT 3A-4a 0 3.5.2-3 Such limitations are not required to be detailed in the Technical Specifications. These surveillance requirements are typically performed during plant shutdown, however, if for instance, an actual signal is generated while operating, results should be useable even though the plant is not " shutdown." Similarly, if testing would be required to complete scme repair or modification made while operating, a shutdown should not be required. Therefore, the CTS wording is consistent with the NUREG-14?1 Bases wording that states: The 18 month Frequency is based on the need to perform these Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surve111ances were performed with the reactor at power. Moving these details to the Bases is consistent with other requirements specified in NUREG-1431, Rev. 1. This change moves to the Bases details 5 that are not necessary to provide operational safety while retaining in technical specifications the basic requirements for maintaining OPERABIL7TY. INSERT 3A-4b 0 3.5.2-5 Plant procedures governing the restoration of equipment after maintenance specify the requirements for determining the appropriate post maintenance testing. Any time the Operability of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate Operability of the system or component. As such, the requirement to perform a flow balance test after modifications that alter ECCS subsystem flow characteristics is not required to be in the TS to provide adequate protection of the public health cnd safety. This requirement has been moved to the [U]SAR (for Callaway, Diablo Canyon, and Wolf Creek) or TRM (Commanche Peak). These licensee controlled documents containing the moved requirements will be maintained using the provisions of 10 CFR 50.59. Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner. INSERT 3A-4c 0 3.5.2-6 i The requirements of ITS LC0 3.5.2 and the associated Surveillance Requirements are adequate to ensure the ECCS are maintained OPERABLE. As a result, the methods of performing Surveillance are not necessary to ensure the ECCS can perform their intended safety function and the details are not required to be in the TS to provide adequate protection of the public health and safety. The ITS Bases containing the moved requirements will be maintained using the provisions of 10 CFR 50.59, as required by the Bases Control Program described in ITS Section 5.5.14. Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner.
ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.5.2-8 APPLICABILITY: WC, CA , REQUEST: DOC 2-19 LG CTS 4.5.2.e The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS. Comment: Please revise the DOC to include additional justification as to why this detail l is not necessary in the ITS. FLOG RESPONSE: DOC 2-19 LG has been revised to provide additional justification for the proposed change by adding the following information:
"Such limitations are not required to be detailed in the Technical Specifications. These surveillance requirements are typically performed during plant shutdown, however, if for instance, an actual signal is generated while operating, results should be useable even though the plant is not " shutdown." Similarly, if testing would be required to complete some repair or modification made while operating, a shutdown should not be required. Therefore, the CTS wording is consistent with the NUREG-1431 Bases wording that states: The 18 month Frequency is based on the need to perform these Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Moving these details to the Bases is consistent with other requirements specified in NUREG-1431, Rev.1. This change moves to the Bases details that are not necessary to provide operational safety while retaining in Technical Specifications the basic requirements for maintaining OPERABILITY."
ATTACHED PAGES: Encl. 3A 4 i < I l l l l
CHANGE MREER MiE DESCRIPTION
, in the improved TS: however, the requirement for an Inservice Testing (IST) Program is moved to Section 5.5.8 of the improved TS. The IST Program is referenced directly for the frequency of testing. h y 34}.4 4 o ss.2-4]
2 13 TR 3 The CTS allowance, which permits the ECCS throttle valves I to be declared OPERABLE without verifying ECCS throttle ! valve stop position for 4 hours following valve stroke testing or maintenance is deleted from the current TS consistent with NUREG 1431 Rev 1. The ECCS throttle valves i are manual valves and plant procedures governing post-l maintenance test requirements specify verification of L correct throttle position prior to declaring the valves l OPERABLE. Explicit post maintenance TS surveillance
)
requirements have been deleted because these requirements ' are adequately addressed by administrative post-l maintenance programs. l 2 14 A Not applicable to WCGS. See Conversion Comparison Table l (Enclosure 38) . 2 15 LG The surveillance requirement for the flow balance test l following ECCS modifications is oved to a licensee controlled document. This requirement is not included in NUREG 1431 Rev. 1. QiasERf 34 -4h 4 3.s.2.-s l 2 16 LG The specific means by which the ECCS piping is assured to be full of water is moved to the Bases. This level of detail is not included in the ISTS and is consistent with ' the kind of information contained in the Bases.
. @ T M -1o2.s.2.-51 2 17 A Adds the phrase "that is not locked. sealed, or otherwise secured in position" with regard to which valves require j actuation testing. This change is merely a clarification. !
Valves that are secured in place are secured in the i position required to meet their safety function. The l actuation testing ensures that valves can move to the position that meets their safety function. If the valves are secured in the position that meets their safety ' function, no testing is necessary. 2 18 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 2 19 LG Consistent with NUREG 1431 Rev. 1. this change moves the requirement that the 18 month verification of automatic ' ECCS valve actuation and ECCS pump actuation be performed during shutdown to the Bases. gitT]JA-QQ3.5.2-el
- WCGS-Description ofChanges to CTS.tM.5 4 5/1SM7
INSERT 3A-4a 0 3.5.2-8 Such limitations are not required to be detailed in the Technical Specifications. These surveillance requirements are typically performed during plant shutdown, however, if for instance, an actual signal is generated while operating, results should be useable even though the ! plant is not " shutdown." Similarly, if testing would be required to complete some repair or modification made while operating, a shutdown should not be required. Therefore, the CTS wording is consistent with l the NUREG-1431 Bases wording that states: The 18 month Frequency is j l' based on the need to perform these Surve111ances under the conditions j that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Moving these details to the Bases is consistent with other requirements specified in NUREG-1431, Rev. 1. This change moves to the Bases details ! that are not necessary to provide operational safety while retaining in technical specifications the basic requirements for maintaining OPERABILITY. t INSERT 3A-4b 0 3.5.2-5 L Plant procedures governing the restoration of equipment after t maintenance specify the requirements for determining the appropriate post maintenance testing. Any time the Operability of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate Operability of the system or component. As such, the requirement to perform a flow balance test after modifications that alter ECCS subsystem flow characteristics is not required to be in the TS to provide adequate protection of the public health and safety. This requirement has been moved to the [U]SAR (for Callaway. Diablo Canyon. and Wolf Creek) or TRM (Commanche Peak). These licensee controlled documents containing the moved requirements will be maintained using the provisions of 10 CFR 50.59. Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner. INSERT 3A-4c 0 3.5.2-6 j The requirements of ITS 1.C0 3.5.2 and the associated Surveillance Requirements are adequete to ensure the ECCS are maintained OPERABLE. As a result, the methods of performing Surveillance are not necessary to ensure the ECCS can perform their intended safety function'and the details are not required to be ir. the TS to provide adequate protection of the public health and safety. The ITS Bases containing the moved requirements will be maintained using the provisions of 10 CFR 50.59, as . required by the Bases Control Program described in ITS'Section 5.5.14. I Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner. l
l l l j ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: O 3.5.3-1 APPLlLABILITY: DC, CP, WC, CA REQUEST: DOC 3-01 LG CTS LCO 3.5.3 ITS LCO 3.5.3
- The referenced DOC describes the change to the CTS but does not provide any justification for making the change other than that it is consistent with the STS.
Comment: Please revise the DOC to inc!ude additional justification as to why this detail ! l is not necessary in the ITS. FLOG RESPONSE: DOC 3-01-LG has been revised to provide additional justification for the proposed change by adding the following information:
"The proposed change is consistent with NUMARC 93-03, " Writer's Guide for the Restructured Technical Specification" and the philosophy of NUREG-1431 in which the LCO describes as simply as possible the lowest functional capability of the system and relegates the details of what constitutes an OPERABLE system to the Bases. Therefore, the details of what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the Bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY."
ATTACHED PAGES: Encl. 3A 5
CHANGE NLABER EiHC DESCRIPTION 3 01 LG Consistent with NUREG-1431 Rev.1, the LCO is revised to replace the word " subsystem" with the word " train" and the descriptive information in the LCO is moved to the BASES. Whereas there is no technical change associated with the replacement of the term " subsystem," " train" better describes that all parts of the required system ( e.g., piping, instruments, controls etc..) must be operable to s the required safety functions. terr 3A -520 l @ J.5.s -# J 3 02 LS-4 Consistent with NUREG-1431 Rev 1, the low temperature overpressure protection limitation on ECCS pumps and related surveillance are moved to improved TS 3.4.12. The prescriptive wording related to pump operability is changed to wording specifically addressing the pumps' capability to inject into the RCS. This change is less restrictive on the configuration of the centrifugal charging and safety injection pumps but is acceptable because it is consistent with the cold overpressure-analysis requirements and still precludes flow to the RCS. 3-03 di 5 fC s h - R 1 / - ed t . The completion time for COLD SHUTDOWN ' I due to CCP inoperability is increased from 20 hours to 24 / I hours. This time is reasonable based on operating I I experience to reach MODE 5 in an orderly manner, without challenging plant systems or operators, and is consistent I with other shutdown action Completion Times to reach j i k5 from MODE 4. susr=AT J4-ob -
~ g s,s.3-7.l 3 04 LG Rev. 1, the ION b fConsistent un NUREG-14 termi y is revis . The requi to re e at lea one ECCS s stem is re to "i atel f tiate acti to restore" RML sub em. h both L AHR pumps heat exc rs inope e, i uld be unwise require t lant to o , where on1 vailable removal tem i RHR. refor ,
appropria action i o ini e measur o re re {oneECCS subsyst and to ntinue t ction ntil the s stem is tored OPERABLE tus. so the alt ate req t f RWt canno res ed) to maintain T < 35 y use of a rnate at remo ) method s des ptive info ion a is mov o the BAS . T ransition to 3 already ohibi in; this se rio by the E speci cation r MOD , 2, , and . tdsEs.T p$r.se.
- 9 3.5.3 4 }
3 05 TR 2 Consistent with NUREG 1431 Rev.1, the requirement to i submit a Special Report within 90 days of an ECCS ! actuation and injection event is deleted. This change is i WCGS-Description of Changes to CTS 3/4.5 5 5/1587 1
INSERT 3A-Sa 0 3.5.3-1
. The proposed change is consistent with NUMARC 93-03. " Writer's Guide for the Restructured Technical Specifications" and the philosophy of NUREG-1431 in which the. LC0 describes as simply as possible the lowest functional capability of the system and relegates the details of what .
I constitutes an OPERABLE system to the Bases. Therefore, the details of what constitutes an OPERABLE subsystem (train) such as required pumps, heat exchangers and flow paths, are more appropriately discussed in the l- bases than in the LCO. These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore, the proposed change moves to the Bases details that are not necessary to provide operational safety while retaining in the technical
. specifications the basic requirements for maintaining OPERABILITY.
l t INSERT 3A-5b 0 3.5.3-2 Due to stable conditions associated with operation in MODE 4. the probability of occurrence of a Design Basis Accident is low. As a result, the ECCS operational requirements are reduced with only one train of the ECCS CCP Subsystem required to be OPERABLE. The required action if the CCP Subsystem is inoperable is to proceed to cold shutdown. INSERT 3A-Sc 0 3.5.3-3 CTS 3.5.3 Action b provides an alternate requirement (if RHR cannot be ! restored) to maintain T. , < 350*F by use of alternate heat removal methods. These details are moved to the ITS Bases. These details are not necessary to ensure the ECCS are OPERABLE. The requirements of ITS
~
, 3.5.3 LC0 and Conditions are adequate for ensuring that the ECCS are l OPERABLE. These details are not necessary to ensure the ECCS can perform their intended safety function. As such, these details are not required to be in the TS to provide adequate protection of the public health and safety. Moving these details maintains consistency with NUREG-1431. Any change to these details will be made in accordance with 10 CFR 50.59 and the Bases Control Program described in ITS Section 5.5.14. l l
- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - \
l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.5.3-2 APPLICABILITY: DC, CP, WC, CA l REQUEST: DOC 3-03 LS-5 ' CTS 3.5.3 Action a { ITS 3.5.3 Actions A & C ' DOC 3-03 LS-5 discussed two distinct changes. The first change involves movement of j the descriptive information to the Bases. The second change is an increase in the completion time to reach Mode 5 from 20 to 24 hours. Comment: The first change, movement of the descriptive information to the Bases, should be separated out and justified as an "LG" change, consistent with other similar changes in this section. The increase in the completion time to reach Mode 5 from 20 to 24 hours is correctly justified as an "LS" change and the justification provided in DOC 3-03 LS-5 is acceptable. FLOG RESPONSE: DOC 3-03 LS-5 has been separated into two DOCS (DOC 3-03-LS and DOC 3-13-LG). DOC 3-03 LS-5 has been revised to address only the increase in completion time. DOC 3-03 LS-5 has been enhanced to include additionaljustification. New DOC 3-13 LG has been created to . address movement of information to the Bases. l ATTACHED PAGES: l 1 Encl. 2 5-7 Encl.3A 5, 6 ! Encl. 3B 4, 5 I I
FMFRGENCY CORE COOLING SYSTEMS 3/15.3 ECCS SUBSYSTEMS - T_ <350*F
~ ~ LIMITING CONDITION FOR OPERATION 3.5.3 As a rrgimum, one ECCS r';rt- xx,.?rd !S ft::' ;ftra"I@shall be l OPERABLE.W $TQ4IZD3M
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'.- ; ; pr-- ^' M "er ; ^. Mr S *^!e', restore at least one gg.g3433p ECCS subsystem to OPERA status within 1 hour or be in COLD huwsdt '
SHUTDOWN within the ne hours.
- b. Wrth no ECCS(residual heat removal (RHR) subsystem OPERABLE, i: _- --- -' M i: :_ : r; -' : Fir S P" 5::t m '. ..; ; [3 44'4R ~j
' "' " -"-- 6mmediately initists action t@ restore at least one ECCS "" Q3.5,3.3l 4
.3.14. A H subsystem to OPERABLE status.
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b""" ECCS mode of operation. WOLF CREEK UNIT 1 3/4 5-7 l Markup ofCTS 3N.S SM$87
7 CHANGE NLMBER H2iG DESCRIPTION 3 01 LG Consistent with NUREG 1431 Rev. 1 the LC0 is revised to replace the word " subsystem" with the word " train" and the descriptive information in the LC0 is moved to the BASES. Whereas there is no technical change associated with the replacement of the term " subsystem," " train" better describes that all parts of the required system ( e.g., piping, instruments, controls etc..) must be operable to s the required safety functions. 14seKT 3A_ -S20 l @ J.5.3 -s J 3 02 LS-4 Consistent with NUREG 1431 Rev 1, the low temperature overpressure protection limitation on ECCS pumps and related surveillance are moved to improved TS 3.4.12. The prescriptive wording related to pump operability is changed to wording specifically addressing the pumps
- capability to inject into the RCS. This change is less restrictive on the configuration of the centrifugal charging and safety injection pumps but is acceptable because it is consistent with the cold overpressure analysis requirements and still precludes flow to the RCS.
3 03 LS 5 fC Y with - R 1 / . ed . The completion time for COLD SHUTDOWN
' due to CCP inoperability is increased from 20 hours to 24/
hours. This time is reasonable based on operating I experience to reach MODE 5 in an orderly manner, without challenging plant systems or operators, and is consistent I with other shutdown action Completion Times to reach I 5 from MODE 4. NsrA.T ar.sv k - _ -- - -
-~ g s.s.3-zl 3 04- LG Rev. 1, the CTION b (Consistent un NUREG 14 termi gy is revis . The requir t to re e at l lea one ECCS s stem is rev' ed to "i atel f tiate acti to restore" RR sub em. h both itHR pumps heat exc rs inope e, i ld be unwise require t lant to o , where on1 vailable removal tem i RHR. refor ,
appropria action i o ini e measur o re re {oneECCS subsyst and to ntinue t ction ntil the s stem is tored OPERABLE tus, so the alt ate req ement f RHR canno res ed) to maintain T < 35 y use of a rnate at remo method s des ptive info ion a is mov o the ;
- BAS . T ransition to E3 already ohibi in l tcis se rio by the E speci cation r MOD , 2, and . Idssa.T pMe.
- qq 3.5.N }
3 05 TR-2 Consistent with NUREG 1431 Rev.1, the requirement to submit a Special Report within 90 days of an ECCS actuation and injection event is deleted. This change is WCGS-Description ofChanges to CTS 3N.5 5 $ASM
INSERT 3A-Sa 0 3.5.3-1
. The proposed change is consistent with NUMARC 93-03. " Writer's Guide for .
the Restructured Technical Specifications" and the philosophy of NUREG- ) , 1431 in which the LC0 describes as simply as possible the lowest l . functional capability of the system and relegates the details of what j constitutes an OPERABLE system to the Bases. Therefore the details of j what constitutes an OPERABLE subsystem (train) such as required pumps. I heat exchangers and flow paths, are more appropriately discussed in the i bases than in the LCO. .These details are not necessary to ensure ECCS OPERABILITY or that the ECCS can perform its intended safety function. Therefore the proposed change moves to the Bases details that are not necessary,to provide operational safety while retaining in the technical specifications the basic requirements for maintaining OPERABILITY. l INSERT 3A-Sb 0 3.5.3-2 Due to stable conditions associated with operation ir. MODE 4, the i
' probability of occurrence of a Design Basis Accident is low. As a l result, the.ECCS operational requirements are reduced with only one
! train of the ECCS CCP Subsystem required to be OPERABLE. The required
- action if the CCP Subsystem is inoperable is to proceed to cold shutdown.
i INSERT 3A-5c 0 3.5.3-3 ! CTS 3.5.3 Action b provides an alternate requirement (f' ; cannot be restored) to maintain T.,, < 350'F by use of alternate hed re*) val methods. -These details are moved to the ITS Bases. These cails are l not necessary to ensure the ECCS are OPERABLE. The requirements of ITS 3.5.3 LCO and Conditions are adequate for ensuring that the ECCS are OPERABLE. These details are not necessary to ensure the ECCS can ) i perform their t'n. ended Safety function. As such, these details are not required to 04 in the. 3 to provide adequate protection of the public health and safety. Moving these details maintains consistency with NUREG-1431. Any . change to these details will be made in accordance with 10 CFR 50.59 and the Bases Control Program described in ITS Section 5.5.14. l l l
I CHANGE Nl#EER j@C DESCRIPTION acceptable because the requirement to submit a report is sufficiently addressed by the reporting requirements contained-in 10CFR50.73. 3 06 A Consistent with TSTF 90, a note is added to the LC0 that clarifies that an RHR subsystem's ECCS function is operable if it is capable of being manually realigned to the ECCS mode of operation. This is an administrative
- hange to provide clarification.
3 07 H The surveillance frequency to verify a maximin of one charging pump is capable of injecting into the RCS is changed from "at least once per 31 days thereafter", to "at least once per 12 hours thereafter". This change is more restrictive with regard to surveillance interval, however, the SR can be performed using control room indication ard administrative controls. 3 08 A A footnote is idded to CTS SR [4.5.3.1) indicating that the SR to verify the RHR interlock action is not applicable when the RHR suction isolation valves are open ; to satisfy LC0 [3.4.9.3]. This LC0 permits the use of the l RHR suction relief valves to satisfy the low temperature ! overpressure protection requirements in [ MODE 3 with any ; RCS cold leg temperature s 368'F and when the RHR system l is in operation] and H0DE 4. When in this configuration, the RHR suction isolation valves are required to be open. 3 09 H Not applicable to WCGS. See Conversion Comparison Table (Enclosure 38). 3 10 LS 6 Consistent with NUREG-1431 Rev. 1, the requirement to demonstrate ECCS train operability in H00E 4 in SR[4.5.3.1] has been revised to delete the 31 day surveillance to verify the correct position of each valve in the ECCS flow path which is not already locked in place and the 18 month surveillance to verify automatic- _ actuation of ECCS pumps and automatic valves. (id5 EAT 3AM Wo s.o.2 r i 3 11 LG Not applicable to WCGS. See Conversion Comparison Table (Enclosure 3B). 3 12 A The SR to verify that no more than one centrifugal charging pump and no SI pumps are capable of injecting l into the RCS and the SR exception for 4 hours after entering MODE 4 from H00E 3 or until the temperature of one or more RCS cold legs decreases below 325'F, whichever comes first are moved to improved TS SR 3.4.12.1, SR m 3.4.12.2. and LC0 3.4.12 Note 2. his t.Cr tussyt.T sA.@ en 3 g.3 2. } WCGS-Description of Changes to CTS 3N.S 6 $/15/97
~
@-14 ~ A i w SE e.T 3 A -@ ---{ 4 3.9.3 -3 h I
INSERT 3A 6a 0 3.5.3-5
-This change is acceptable because the ECCS operational requirements can be reduced due to the stable conditions associated with operation in MODE 4 and the decreased probability of occurrence of a Design Basis Accident (OBA). ECCS operational requirement reductions mean that certain automatic safety injection (SI) actuation signals are not available. However, in MODE 4, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
INSERT 3A-6b 0 3.5.3-2 3-13 LG CTS LCO 3.5.3 Action a. terminology is revised and the descriptive information moved to the ITS Bases. These details are not necessary to ensure the ECCS Required Actions are met. The requirements of ITS 3.5.3 LC0 and Conditons are adequate for ensuring the ECCS are OPERABLE. These details are not necessary to ensure the ECCS can perform their intended safety function. As such, these details are not required to be in the TS to provide adequate protection of the public health and sa'fety. Moving these details maintains consistency with NUREG-1431. Any change to these details will be made in accordance with 10 CFR 50.59 and the Bases Control Program descriuea in ITS Section 5.5.14. INSERT 3A-6c 0 3.5.3-3 3-14 A CTS LCO 3.5.3 Action b. provides, with no ECCS RHR subsystem OPERABLE. the option to either restore at least one ECCS RHR subsystem to OPERABLE status or to maintain the RCS T.,, < 350*F by use of alternate heat removal methods. Condition A of ITS LC0 3.5.3 requires that with no ECCS RHR subsystems OPERABLE that immediate action be initiated to restore an ! ECCS RHR subsystem to OPERABLE status. While the CTS does not specify a time frame to initiate action to restore one ECCS RHR subsystem, the current operational philosophy is that this action is initiated immediately. The Completion j Time of "immediately" to initiate actions that would restore l at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling ; i capacity. Revising the CTS Action to immediately initiate action is considered an administrative change and is consistent with NUREG-1431. E - __ _ _ _ _
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