ML20138E483

Ack Receipt of 970319 Rev of TS Bases Into Plant TS & Informs That Review Found Rev Acceptable

ML20138E483
Person / Time
Site:	Wolf Creek
Issue date:	04/24/1997
From:	Stone J NRC (Affiliation Not Assigned)
To:	Maynard O WOLF CREEK NUCLEAR OPERATING CORP.
References
TAC-M98202, NUDOCS 9705020243
Download: ML20138E483 (6)

Text

_

April 24, 1997 l l

l e Mr. Otto L. Maynard  !

! President and Chief Executive Officer  !

i Wolf Creek Nuclear Operating Corporation l 1 Post Office Box 411 l 1

Burlington, Kansas 66839 l 2

1

SUBJECT:

WOLF CREEK GENERATING STATION - TECHNICAL SPECIFICATION BASES CHANGE (TAC N0. M98202)

Dear Mr. Maynard:

The staff has incorporated the revision of the technical specification Bases l provided by your letter of March 19, 1997, into the Wolf Creek Generating Station Technical Specifications.

The staff has reviewed the changes and find the revisions to the associated i technical specification Bases to be acceptable. The overleaf pages are i

.j provided to maintain document completeness.  !

Sincerely, J

Original Signed by Kristine Thomas for f

i James C. Stone, Senior Project Manager ,

1 Project Directorate IV-2  !

j Division of Reactor Projects III/IV  :

j Office of Nuclear Reactor Regulation j l

Docket No. 50-482 DISTRIBUTION: I LDocket-GHill (2)

Enclosure:

Bases Pages PUBLIC CGrimes PDIV-2 Reading WJohnson, Region IV cc w/ enc 1: See next page JRoe LHurley, Region IV  ;

EAdensam JKilcrease, Region IV l WBateman JCalvo JStone EPeyton 0GC ACRS Document Name: WC98202.LTR OFC PDIV-2 PDIV-2 EELB yyfd NAME EPEpik 3Y oYe" JCalvo

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DATE 4fLt/97 4/al/97 4/'U/97 - g g\

/n -41 0FFICIAL RECORD COPY p c2003 NRC HLE CEMEt COPY 9705020243 970424 PDR ADOCK 05000482 P PDR l

Mr. Otto L. Maynard -2_ April 24, 1997 i i cc w/ enc 1:

Jay Silberg, Esq. Chief Operating Officer

• Shaw, Pittman, Potts & Trowbridge Wolf Creek Nuclear Operating Corporation '

2300 N Street, NW P. O. Box 411 Washington, D.C. 20037 Burlington, Kansas 66839 Regional Administrator, Region IV Supervisor Licensing U.S. Nuclear Regulatory Commission Wolf Creek Nuclear Operating Corporation l 611 Ryan Plaza Drive, Suite 1000 P.O. Box 411 l Arlington, Texas 76011 Burlington, Kansas 66839 l Senior Resident Inspector U.S. Nuclear Regulatory Commission i U.S. Nuclear Regulatory Commission Resident Inspectors Office i P. O. Box 311 8201 NRC Road

Burlington, Kansas 66839 Steedman, Missouri 65077-1032 i

Chief Engineer  ! ' Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Road ] Topeka, Kansas 66604-4027 Office of the Governor State of Kansas j Topeka, Kansas 66612 Attorney General Judicial Center 301 S.W. 10th 2nd Floor

Topeka, Kansas 66612 l County Clerk Coffey County Courthouse 3 Burlington, Kansas 66839 Public Health Physicist Bureau of Air & Radiation Division of Environment 3

Kansas Department of Health j and Environment Forbes Field Building 283 Topeka, Kansas 66620 i i i 4 a

l 3/4.8 ELECTRICAL POWER SYSTEMS RASES l k 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES. D.C. CMRCES. AND ONSITE p0WER i DISTRIBUT ON ! The OPERABILITY of the A.C. and D.C power sources and associated distri-l bution systems during operation ensures taat sufficient power will be avail-t able to supply the safety-related equipment required for: (1) the safe shut-

!                      down of the facility, and (2) the mitigation and control of accident condi-tions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements j                       of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The ACTION requirements specified for the levels of degradation of the i power sources provide restriction upon continued facility operation commen-l surate with the level of degradation. The OPERASILITY of the power sources are consistent with the initial condition assumptions of the safety analyses j and are based upon maintaining at least one redundant set of onsite A.C. ano D.C. power sources and associated distribution systems OPERA 8LE during acci-l dent conditions coincident with an assumed loss-of-offsite power and single  ! failure of the other onsite A.C. source. The A.C. source and D.C. source l allowable out-of-service times are based on Regulatory Guide 1.93, I

                      ' Availability of Electrical Power Sources", December 1974. When one diesel I

j generator is inoperable, there is an additional ACTION requirement to verify ' that all required systems, subsystems, trains, components and devices, that depend on the remaining 0PERA8LE diesel generator as a source of emergency .

'                    power, are also OPERABLE, and that the stear 1 riven auxiliary feedwater pump f              is OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is j

inoperable. The tem verify as used in this context means to administrative 1y check by examining logs or other infomation to detemine if certain i components are out-of-service for maintenance or other reasons. It does not i mean to perform the Surveillance Requirements needed to demonstrate the l OPERA 8ILITY of the component. l The OPERABILITY of the minimum specified A.C. and D.C. power sources and i associated distribution systems during shutdown and refueling ensures that: ! (1) the facility can be maintained in the shutdown or refueling condition i for extended time periods, and (2) sufficient instrumentation and control j capability are available for monitoring and maintaining the unit status. 1 When detemining compliance with action statement requirements, addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered to be a positive reactivity change. l The surveillance requirements of Technical Specification 3/4.8.1 are based upon, in part, the guidance of Generic Letter M-01, " Removal of ! Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators from Plant Technical Specifications," Generic Letter 93-05, 'Line-Item Technical Specifications Improvements to Reduce surveillance Requirements i for Testing During Power Operation," Regulatory Guide 1.9, " Selection, Design, j Qualification, and Testing of Emergency Diesel Generator Units Used as Class

 ~

IE Onsite Electrical Power Systems at Nuclear Power Plants,' Revision 3, and i

 !                   WOLF CREEK - UNIT 1                                      8 3/4 8-1   Amendment No. 0,50,55,101 j                                                                                           November 22, 1993 i

l i l

l ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES. D.C. SOURCES. AND ONSITE POWER DISTRIBUTION (Continued) NUREG-1431, " Standard Technical Specifications - Westinghouse Plants." Also, the guidance of NUMARC 87-00, " Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," Revision 1, and Regulatory Guide 1.160 has been adopted to formulate a comprehensive Emergency Diesel Generator Reliability Program. Technical Specification 3.8.1.1, Action b and c, require, in part, the demonstration 01 3e operability of the remaining operable emergency diesel generator by perfornJng Technical Specification 4.8.1.1.2a.4. This test is required to be completed regardless of when the inoperable emergency diesel generator is restored to operable status unless the emergency dieml generator was declared inoperable to do preplanned preventative maintenance, testing, or maintenance to correct a condition which, if left uncorrected, would not affect the operability of the emergency diesel generator. The requirement to test the remaining operable emergency diesel generator when one emergency diesel generator is inoperable is limited to those situations where the cause for inoperability cannot be conclusively demonstrated in order to preclude the potential for comon mode failures. The test is not required to be accomplished if the emergency diesel generator was declared inoperable due to an inoperable support system or an independently testable component. When such a test is required, it is required to be performed within 24 hours for ACTION b and within 8 hours for ACTION c of having determined that the emergency diesel generator is ineperable. Technical Specification 4.8.1.1.2a.4 is considered to be a " Start Test" as described in Regulatory Guide 1.9, Revision 3. A " Start Test" is performed to demonstrate proper startup from standby conditions and to verify that the required design voltage and frequency is attained. For these tests, Regulatory Guide 1.9, Revision 3, recommends that the emergency diesel generators be slow started and allowed to reach rated speed on a prescribed schedule that is selected to minimize stress and wear. Regulatory Guide 1.9, Revision 3, considers Technical Specification 4.8.1.1.2a.5 to be a " Load-Run Test". A " Load-Run Test" demonstrates 90 to 100 percent (5580 to 6201 kilowatts) of the continuous rating (6201 kilowatts) of the emergency diesel generator for an interval of not less than 1 hour and until temperature equilibrium has been attained. This test may be accomplished by synchronizing the generator with offsite power and the loading and unloading of a diesel generator during this test should be gradual and 4 based on a prescribed schedule that is selected to minimize stress and wear on the diesel generator. Regulatory Guide 1.9, Revision 3, considers Technical Specification 4.8.1.1.2f to be a " Fast-Start Test". A " Fast-Start Test" demonstrates that l each emergency diesel generator starts from standby conditions. If a plant normally has in operation keep warm systems designed to maintain lube oil and jacket water cooling at certain temperatures or prelubrication systems or both, this would constitute normal standby conditions for that plant. Verification that the emergency diesel generator reaches required voltage and , frequency within acceptable limits and time is also required.  ; WOLF CREEK - UNIT 1 B 3/4 8-2 Amendment No. 101 April 24, 1997 i

o 3/4.2 POWER DISTRIBUTION LIMITS

   /

BASES The specifications of this section provide assurance of fuel integrity during Condition I (Nomal Operation) and II (Incidents of Moderate frequency)  ; events by: (a) maintaining the minimus DNBR in the core greater than or equal to the DNBR design limit specified in the CORE OPERATING LIMITS REPORT (COLR) s during normal operation and in short-term transients, and (b) limiting the fission gas release, fuel pellet temperature, and cladding mechanical proper-ties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial i conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: F,(Z) Heat Flux Hot Channel Factor, is defined as the local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, Fj Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure tha.t the F Z) and F5 limits are not exceeded during either nomal operation or in T(h xenon redistribution following power changes. The AFD limits have been adjusted for measurement uncertainty. r Provisions for monitoring the AFD on an automatic basis are derived from  : the plant process computer through the AFD Monitor Alarm. The computer deter- l sines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is greater than 50% of RATED THERMAL POWER. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CPA"El FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR t The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. 1 WOLF CREEK - UNIT 1 B 3/4 2-1 Amendment No. he,92

POWER DISTRIBUTION LIMITS i BASES i 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHAlpY RISL HOT CHANNEL FACTOR (Continted) l Each of these is measurable but will normally only be determined ' periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic i surveillance is sufficient to insure that the limits are maintained provided: 2 l' a. Control rods in a single group move together with no individual rod insertion differing by more than 12 steps, indicated, from the

group demand position,

b. Control rod groups are sequenced with overlapping groups as

described in Specification 3.1.3.6,

)

c. The control rod insertion limits of Specification 3.1.3.6 are l

. maintained, and j j d. The axial power distribution, expresseJ in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. 1 i F5 will be maintained within its limits provided Conditions a. through

d. above are maintained. The limits on the nuclear enthalpy rise hot channel factor, F5 are specified in the COLR.

4 Fo (Z) and Fj are measured periodically to provide assurance that they remain within their limits. A peaking margin calculation is performed, when necessary, to provide the basis for reducing THERMAL POWER or for reducing the { I width of the AFD limits. The hot channel factor f(Z) is measured periodically and increased by a cycle and height d,ependent factor, W(Z), to provide assurance that the limit of Fa(Z) is met. W(Z) accounts for the effects of normal operation transients and is determined from expected power control maneuvers over the full range of burnup conditions in the core. The  ; W(Z) functions are specified in the Core Operating Limits Report. 1 3/4.2.4 OVADRANT POWER TILT RATI_Q The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are'made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. WOLF CREEK - UNIT 1 B 3/4 2-2 Amendment No. 1,51,51,92 l April 24, 1997}}