Text

License Amendment Request 187, Conforming Technical Specification Changes for Use of Westinghouse Vantage+ Fuel. Attachments 1 - 4 Which Includes Description of Proposed Changes, Safety Evaluation
	ML022200535
Person / Time
Site:	Kewaunee
Issue date:	07/26/2002
From:	Warner M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC-02-067
	Download: ML022200535 (160)
	v • d • e

ATTACHMENT 1 Letter from M. E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 187 Description of Proposed Changes Safety Evaluation Significant Hazards Determination Environmental Consideration

Docket 50-305 NRC-02-067 July 26, 2002, Page 1 Introduction Nuclear Management Company, LLC (NMC) intends to operate its Kewaunee Nuclear Power Plant (KNPP) reactor using Westinghouse 422 VANTAGE+ nuclear fuel with PERFORMANCE+

features (referred to subsequently as 422V+), commencing with the Cycle 26 core refueling presently scheduled for April of 2003.

The 422V+ fuel design proposed by this request is essentially the same as the fuel that the Nuclear Regulatory Commission (NRC) approved (Reference 1) for use at Point Beach Nuclear Plant (PBNP) as requested by Wisconsin Electric (References 2 and 3). Design features of the 422V+ fuel were described in detail by a Westinghouse topical report, WCAP-12610-P-A (Reference 4), which was referenced by the PBNP request. The NRC subsequently approved the PBNP request and issued amendments 193 and 198 for facility operating licenses DPR-24 and DPR-27, respectively.

Prior to the use of 422V+ fuel for full reloads, it is necessary to make changes to KNPP Technical Specifications (TS). Those changes are shown in attachments to this request. NMC requests NRC permission for the TS amendments and provides associated USAR changes as information for NRC use in evaluating this request.

NMC worked with Westinghouse to review the KNPP design and evaluate use of 422V+ fuel. The reanalysis was performed using NRC approved, Westinghouse computer codes and methods, thereby establishing a new Record of Analysis which is documented in the Technical Design Basis for the change (Attachment 4) and in the USAR markups (Attachment 5). The analyses were performed assuming a KNPP reactor power level of 1772 MWt rather than the currently licensed level of 1650 MWt. The use of the higher power level in the technical design basis calculations adds conservatism to analytical results; however, operation at the higher level is not requested by this proposal.

The Westinghouse Report serves as a reference safety evaluation and analysis report for the region by-region reload transition from the KNPP Cycle 25 reactor core to subsequent cores containing a mix of 422V+ and earlier designs, and ultimately to a homogeneous core loaded solely with 422V+

fuel. Thus, it will be a reference for future KNPP Reload Safety Evaluations (RSE). Each RSE will be conducted to show cycle-specific conformance within the parameters established by the RTSR or require cycle-specific analyses or evaluations to ensure safe operation of each core design, including Cycle 26.

NRC approval of this TS amendment is requested by February 28, 2003, for implementation during the Cycle 26 refueling outage.

Docket 50-305 NRC-02-067 July 26, 2002, Page 2 Description of Changes to KNPP Technical Specifications (TS)

The following is a list of changes to the Technical Specification, a description of each change, and the reason that the change is being made:

TS Section Description of Chan~es Reason L.p Changed to Dose Equivalent 1-131 Revised methodology and definition and dose conversion assumptions that are derived from factors TID-14844 to ICRP-30 consistent with radiological consequences analyses.

2.1.b Added WRB-1 Correlation Change reflects the new DNB Basis TS 2.1 correlation used in analyses for Westinghouse 422 V+ fuel.

Basis TS 2.2 Modified PORVs basis Clarified the design basis for the Pressurizer PORVs by indicating the design basis events and the pressure value acceptance criterion which is applicable to the PORVs.

2.3.a.3.A Changed Overtemperature AT term Changed to be consistent with descriptions analysis assumptions.

2.3.a.3.B Changed Overpower AT term Changed to be consistent with descriptions analysis assumptions. There is no f(AI) penalty for this function.

3.1.c Changed specific activity values, Changes made for consistency with Basis TS 3.1.c replaced 10 CFR 100 with 10 CFR ICRP-30 and radiological Basis TS 3.4.c 50.67, and deleted Figure 3.1-3, the consequence analyses.

Dose Equivalent 1-131 figure.

Basis TS 3.1.a Changed DNBR value Change is associated with 422 V+

fuel.

3.4.d Changed the units "ptCi/cc" to Changes made for consistency with "RCi/gram ICRP-30 and radiological consequence analyses.

Basis TS 3.4 Changed Maximum full-power steam Change made to be consistent with flow and AFW operable description, analyses.

3.10.b Changed the actions for FoN(Z), FAHN, These actions were changed to be 3.10.b.7 and FQEQ(Z) consistent with the FQ surveillance Basis TS 3.10.b methodology and 422 V+ fuel Basis TS 3.10.b change.

3.10.b.7 through Changed the surveillance and actions These changes were changed to be 3.10.b.13 for axial flux difference consistent with the relaxed axial Basis TS 3.10.b offset control (RAOC) methodology and 422 V+ fuel change.

3.1 0.m Changed the surveillances for reactor These changes are made to be Basis TS 3.10.k coolant flow and bases for pressure, consistent with analysis assumptions.

Basis TS 3.10.1 temperature and flow.

Basis TS 3.10.m 6.9 Added reference Reference is added for COLR methods documentation.

Docket 50-305 NRC-02-067 July 26, 2002, Page 3 TS Section Description of Changes Reason 2.6 Added new FoN(Z),

FAHN, and Maximum FQ and FmN are generated FQEQ(Z) limits for 422 V+ fuel that yield acceptable results based upon the safety analysis limits. For F0 see Section 3.5.

For F N see Sections 3.6 and 4.2.

2.6 Replaced the V(z) function with The W(z) function is used for W(z) function.

consistency with the FQ Surveillance Methodology.

2.7 Increased the part power multiplier This change was made for for FAH.

consistency with analysis inputs.

2.9 Changed the OTAT function.

This change was made to reflect the impact of the transition to 422 V+

fuel and for consistency to the Non LOCA Analysis.

2.10 Changed the OPAT function.

This change was made to reflect the impact of the transition to 422 V+

fuel and for consistency to the Non LOCA Analysis.

2.11 Revised the DNB Parameters (flow, The flow change is made to be pressure and temperature) consistent with the Westinghouse Methodology for RCS flow measurement.

The pressure and temperature change is made to reflect KNPP specific uncertainties, procedures, and calibration practices in generating the indicated values.

2.12 Revised Refueling Boron Boron concentration change was Concentration Value from 2200 made for consistency with the ppm to 2250 ppm accident analyses.

Figure 1 Revised the Core Safety Limits This change was made to reflect the Curve impact of the transition to 422 V+

fuel and for consistency to the Non LOCA Analysis.

Figure 2 Revised Required Shutdown The SDM change was made to Margin (SDM) provide additional fuel management flexibility and is consistent with accident analysis.

Figure 3 Revised Hot Channel Factor The K(z) curve was changed for Normalized Operating Envelope consistency with the LOCA analysis.

K(z)

Figure 5 Replaced the V(z) figure with the W(z) figures.

The W(z) function is used for consistency with the FQ Surveillance Methodology.

Docket 50-305 NRC-02-067 July 26, 2002, Page 4 Safety Evaluation for Proposed Change to KNPP TS Changes proposed by this request modify the KNPP TS to allow use of full reloads of Westinghouse 422V+ fuel beginning with Cycle 26. Discharged fuel will be replaced by 422V+ fuel during this and subsequent refueling operations, until the core contains only 422V+ fuel.

Compared with current Framatome/ANP design, 422V+ fuel has a slightly smaller fuel rod diameter.

Other minor differences exist and are presented in Table 2-1 of the Technical Design Basis for the Transition to 422V+ Fuel in Attachment 4 (heretofore referred as Westinghouse Report). While the current fuel uses a zirconium alloy known as Zircaloy for fuel clad and for structural elements of the fuel assembly, 422V+ fuel uses a substantially similar zirconium alloy marketed under the Westinghouse ZIRLO trademark. This new name requires changes to TS and USAR references that do not presently cite ZIRLO as an acceptable material. Westinghouse 422V+ fuel with ZIRLO is mechanically, hydraulically, and chemically compatible with the KNPP reactor design and the Framatome/ANP fuel currently loaded in its core.

During fuel transition cycles, KNPP cores will contain combinations of partially burned Framatome/ANP fuel and 422V+ fuel, and fresh 422V+ fuel. The Westinghouse Report discusses fuel design changes and describes the safety analyses performed to bound core conditions containing any combination of these fuel designs, including a full core load of 422V+ fuel. The effect of 422V+

on new-fuel and spent-fuel handling and storage was analyzed and remains acceptable. For Cycle 26 and subsequent core loads, cycle-specific RSE will be performed using Westinghouse standard reload methodology (Reference 6) to verify that safety analaysis results satisfy their respective acceptance criteria.

Note that certain of the analyses performed for this request assumed a reactor power of 1772 MWt.

No power uprate request is made at this time. Conclusions reached using the 1772 MWt assumption bound results using the current power level of 1650 MWt. Reactor trip set-points calculated using the higher power level also bound safe operation at the currently licensed power level.

Westinghouse reanalyzed many of the existing safety analyses as documented in Attachment 4, to revise or confirm assumptions and inputs to ensure their accuracy and that of the KNPP licensing basis. References 1-17, 1-18, and 1-19 document the final assumptions and input parameters used in the analyses. However, the individual calculation notes (Reference 5, Table 5.1-8) should be used as the final verification of all assumptions and parameters.

The Westinghouse Report documents extensive analyses that led to the findings summarized below.

Together, these findings conclude that plant operation using 422V+ fuel is safe, either in combination with various mixes of existing fuel of other designs, or as the sole core-resident fuel.

" 422V+ fuel is mechanically compatible with current Framatome/ANP fuel assembly design, control rods, flux detectors inserted in instrumentation tubes, fuel handling equipment, and reactor internals (Chapters 2 and 6).

Evaluation of 422V+ fuel for postulated faulted condition accidents, including LOCA and seismic events, demonstrate that the fuel assembly is structurally adequate to prevent grid-crush by resultant component stresses and grid impact forces (Chapters 2, 5, and 6).

Docket 50-305 NRC-02-067 July 26, 2002, Page 5 "NMC and Westinghouse core design management using Reload Safety Evaluation (RSE) analyses will ensure that changes in nuclear characteristics of 422V+ fuel during the transition period and subsequent cycles will remain within the ranges analyzed in the Westinghouse Report. Alternatively, additional analyses or evaluations will be performed to demonstrate that the plant will be operated in a manner compliant with all safety criteria. (Chapter 3).

"* 422V+ fuel is hydraulically compatible with current Framatome/ANP fuel (Chapters 2 and 4).

"* Core design and safety analyses of 422V+ fuel confirm that parameters and conditions at an uprated 1772 MWt power level remain within licensed safety margins (Chapters 3, 4, 5 and 6).

"* The analyses and evaluations summarized in Attachment 4 provide the design basis for Westinghouse RSE of future reloads using 422V+ fuel. (Chapters 2, 3, 4, 5, and 6).

The NRC has generically approved use of Westinghouse 422V+ fuel (Reference 4), and has generically approved (Reference 7) the use of a limited number of assemblies as lead test assemblies at KNPP. During Cycle 25 (Reference 8), NMC has confirmed that the fuel is compatible with the KNPP reactor and the existing Framatome/ANP fuel. Analysis of transition and equilibrium 422+

fuel in either mixed or homogenous core configurations remains within design basis limitations and safety margins. Thus, operation with 422V+ fuel remains bounded by design basis accident and transient analyses.

Commitment Upon NRC approval of WCAP 12488, Addendum 2, "Revision to Design Criteria," confirmation of meeting the criterion will be submitted.

Significant Hazards Determination for Proposed Change to TS NMC reviewed the proposed change in accordance with provisions of 10 CFR 50.92 and determined that it produces no significant hazards. The proposed change does not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The NRC generically approved Westinghouse 422V+ fuel assemblies for use in reactors substantially similar to KNPP. NMC used 422V+ fuel in the Lead-Test-Assembly Program during Cycle 25, as permitted by existing TS. Empirical data acquired during Cycle 25 confirms that this fuel is both compatible with KNPP reactor design and with the Framatome/ANP fuel currently in use. Reanalysis of postulated KNPP design basis accidents shows that reactor operation with 422V+ fuel remains within design basis limitations and safety margins. All design basis accidents and transients affected by the fuel upgrade were analyzed, and the results documented in the Westinghouse Report provided with this request. These analyses and evaluations show that use of 422V+ fuel is acceptable. The margin to safety is not exceeded in any instance. Pending approval of Addendum 2 to WCAP 12488 revising the current transient stress strain criteria, all design basis acceptance criteria will be satisfied. Changes to the technical specification that remain within the limits of the bounding accident analyses cannot change the probability or consequence of an accident previously evaluated. Thus, nothing in this proposal will cause an increase in the probability or consequence of an accident previously evaluated.

Docket 50-305 NRC-02-067 July 26, 2002, Page 6

2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

Use of the 422V+ fuel is consistent with current plant design bases and does not adversely affect any fission product barrier, nor does it alter the safety function of safety significant systems, structures and components or their roles in accident prevention or mitigation. The operational characteristics of 422V+ fuel are bounded by the safety analyses (Attachment 4). The 422V+

fuel design performs within existing fuel design limits. Thus, this proposal does not create the possibility of a new or different kind of accident.

3) Involve a significant reduction in the margin of safety.

The proposed change does not alter the manner in which Safety Limits, Limiting Safety System Set-points, or Limiting Conditions for Operation are determined. Licensed safety margins are maintained. It conforms to plant design bases, is consistent with current safety analyses, and limits actual plant operation within analyzed and licensed boundaries. Analyses of design basis accidents and transients were performed using a power level greater than that currently licensed, thus rendering more conservative results than required. All safety analysis acceptance criteria are satisfied at this value and all KNPP safety requirements continue to be met. Use of 422V+

fuel as proposed by this amendment request is bounded by these analyses. Thus, changes proposed by this request do not involve a significant reduction in the margin of safety.

Environmental Considerations This proposed amendment involves a change to the Technical Specifications. It does not modify any facility components located within the restricted area, as defined in 10 CFR 20. NMC has determined that the proposed amendment involves no significant hazards considerations and no significant change in the types of effluents that may be released offsite and that there is no significant increase in the individual or cumulative occupational radiation exposure. This proposed amendment accordingly meets the eligibility criteria for categorical exclusion set forth by 10 CFR 51.22(c)(9).

Pursuant thereto, no environmental impact statement or environmental assessment need be prepared in connection with this amendment proposal.

References:

1. USNRC Safety Evaluation Report, "SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO 193 TO FACILITY OPERATING LICENSE NO. DPR-24 AND AMENDMENT NO. 198 TO FACILITY OPERATING LICENSE NO. DPR-27 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301"

2. Wisconsin Electric Company letter NPL 99-0369, "TECHNICAL SPECIFICATION CHANGE REQUEST 210 AMENDMENT TO FACILITY OPERATING LICENSES TO REFLECT REQUIRED CHANGES TO THE TECHNICAL SPECIFICATIONS AS A RESULT OF USING UPGRADED FUEL POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 " from Mark E. Reddemann to Document Control Desk, dated June 22, 1999

Docket 50-305 NRC-02-067 July 26, 2002, Page 7

3. Wisconsin Electric Company letter NPL 99-0731, "Supplement 1 to Technical Specification Change Request 210, Point Beach Nuclear Plant, Units 1 and 2," from Mark E. Reddemann to Document Control Desk, dated December 17, 1999

4. NRC Letter, Ashok Thadani to S.R. Tritch, "Acceptance for Referencing Topical Report WCAP 12610 'VANTAGE + Fuel Assembly Reference Core Report',", July, 1991

5. Westinghouse "Reload Transition Safety Report for the Kewaunee Nuclear Power Plant," dated July 2002

6. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," by S. L.

Davidson, et al, dated July 1985

7. WPSC Letter NRC 91-084 from K.H. Evers to U.S. Nuclear Regulator Commission Document Control Desk, "Core Reloads of Advanced Design Fuel Assemblies", dated June 19, 1991

8. Kewaunee Nuclear Power Plant, Reload Safety Evaluation Cycle 25, November 2001, Nuclear Management Company, LLC, Wisconsin Public Power Corporation, Wisconsin Power & Light

9. LTR-NRC-02-18, Submittal of WCAP-12488-A, Addendum 2/WCAP-14204-A, Addendum 2 of Westinghouse Fuel Criteria Evaluation Process, Revision to Design Criteria", April 26, 2002

ATTACHMENT 2 Letter from M. E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 187 Strike-Out Pages for Technical Specifications including Bases TS ii TS vi TS 1.0-6 TS 2.1-1 TS 2.3-2 through TS 2.3-3 TS 3.1-6 TS 3.4-3 TS 3.10-1 through TS 3.10-5 TS 3.10-9 TS Figure 3.1-3 TS 5.2-1 TS 5.3-1 TS 6.9-4 TS B2.1-1 TS B2.2-1 TS B3.1-8 TS B3.4-1 through TS B3.4-4 TS B3.10-2 through TS B3.10-5 TS B3.10-8

3.2 Chemical and Volume Control System..........................................................

3.2-1 3.3 Engineered Safety Features and Auxiliary Systems......................................

3.3-1 3.3.a Accumulators...........................................................................

3.3-1 3.3.b Emergency Core Cooling System.............................................

3.3-2 3.3.c Containment Cooling Systems..................................................

3.3-4 3.3.d Component Cooling System.....................................................

3.3-6 3.3.e Service W ater System..............................................................

3.3-7 3.4 Steam and Power Conversion System........................................................

3.4-1 3.4.a Main Steam Safety Valves........................................................

3.4-1 3.4.b Auxiliary Feedwater System.....................................................

3.4-2 3.4.c Condensate Storage Tank......................................................

3.4-43=

3.4.d Secondary Activity Limits........................................................

3.4-63 3.5 Instrumentation System.............................................................................

3.5-1 3.6 Containment System.................................................................................

3.6-1 3.7 Auxiliary Electrical Systems........................................................................

3.7-1 3.8 Refueling Operations.................................................................................

3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits................................................

3.10-1 3.10.a Shutdown Reactivity...............................................................

3.10-1 3.10.b Power Distribution Limits........................................................

3.10-1 3.10.c Quadrant Power Tilt Limits....................................................

3.10-43 3.10.d Rod Insertion Limits................................................................

3.10-4 3.10.e Rod Misalignment Limitations.................................................

3.10-5 3.10.f Inoperable Rod Position Indicator Channels........................... 3.10-5 3.10.g Inoperable Rod Limitations.....................................................

3.10-6 3.10.h Rod Drop Time.......................................................................

3.10-6 3.10.i Rod Position Deviation Monitor...............................................

3.10-6 3.10.j Quadrant Power Tilt Monitor...................................................

3.10-6 3.10.k Core Average Temperature....................................................

3.10-6 3.10.1 Reactor Coolant System Pressure..........................................

3.10-6 3.10.m Reactor Coolant Flow..........................................................

3.10-76 3.10.n DNBR Parameters................................................................

3.10-76 3.11 Core Surveillance Instrumentation............................................................

3.11-1 3.12 Control Room Post-Accident Recirculation System..................................

3.12-1 3.14 Shock Suppressors (Snubbers)................................................................

3.14-1 4.0 Surveillance Requirements......................................................................................

4.0-1 4.1 Operational Safety Review.........................................................................

4.1-1 4.2 ASME Code Class In-service Inspection and Testing.................................

4.2-1 4.2.a ASME Code Class 1, 2, 3, and MC Components and Supports..................................................................................

4.2-1 4.2.b Steam Generator Tubes...........................................................

4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection....................................................

4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection....................................................

4.2-3 4.2.b.3 Inspection Frequency.........................................

4.2-4 4.2.b.4 Plugging Limit Criteria.........................................

4.2-5 4.2.b.5 Deleted 4.2.b.6 Deleted 4.2.b.7 Reports...............................................................

4.2-5 4.3 Deleted LAR 187 TS ii 07/26/2002 EAge Section Title

LIST OF FIGURES FIGURE TITLE 2.1-1................. Deleted 3.1-1................. Heatup Limitation Curves Applicable for Periods Up to 33'1]

Effective Full-Power Years 3.1-2................. Cooldown Limitation Curves Applicable for Periods Up to 3311]

Effective Full-Power Years 3.1-3............

D Equivalent 1131 R-oacr, Colant Spocific Acti.ity Limit

%Ir.,

e D -,,..

D,-*÷A. Yl r'h..rr,

D,.,,

-ilst t

w4 w.w.n wD ea le t 3.1-4................. Deleted 3.10-1............... Deleted 3.10-2............... Deleted 3.10-3............... Deleted 3.10-4............... Deleted 3.10-5............... Deleted 3.10-6............... Deleted 4.2-1................. Deleted 5.4-1................. Minimum Required Fuel Assembly Bumup as a Function of Nominal Initial Enrichment to Permit Storage in the Tranfer Canal Note:

[1]

Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

TS vi LAR 187 07/26/2002

p.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 is that concentration of 1-131 (/uCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be as listed and calculated based on dose conversion factors derived from ICRP-30.with tho mo.thoedo*gy established in Tab" o III of TID 18*14, "Calculaieon of Dictanco AFActr for-PoWor and Tost Roactor Sitoc."

DOSE CONVERSION FACTOR J

ISOTOPE 1.0000 1-131 0.03W059 1-132 0.2*4O31692 1-133 0.046"0010 1-134 0.08380293 1-135

q.

CORE OPERATING LIMITS REPORT (COLR)

The COLR is the unit specific document that provides cycle-specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.a.4. Plant operation within these limits is addressed in individual Specifications.

r. SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means (TS 3.10.e), it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

b. In the OPERATING and HOT STANDBY MODES, the fuel and moderator temperatures are changed to the program temperature.

LAR 187 TS 1.0-6 07/26/2002

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE APPLICABILITY Applies to the limiting combination of thermal power, Reactor Coolant System pressure and coolant temperature during the OPERATING and HOT STANDBY MODES.

OBJECTIVE To maintain the integrity of the fuel cladding.

SPECIFICATION

a. The combination of rated power level, coolant pressure, and coolant temperature shall not exceed the limits specified in the COLR. The safety limit is exceeded if the point defined by the combination of Reactor Coolant System average temperature and power level is at any time above the appropriate pressure line.

b. The departure from nucleate boiling ratio (DNBR) shall be maintained _> 1.14 for the HTP DNB correlation and 1.17 for the WRB-1 DNB correlation

c. The peak fuel centerline temperature shall be maintained < 4700 oF LAR 187 07126=2002 TS 2.1-1

3.

Reactor Coolant Temperature A.

Overtemperature AT <ATo [K 0

-K2(T-T')I+rS+K3(P-P')-f(AI)]

1+ T2 S where AT0

=

Indicated AT at RATED POWER, % RATED POWER T

=

A....ag. tempratu--'roReference Average Temperature at RATED POWER, OF T'=

[*]OF P

=

Pressurizer pressure, psig P'

=

[*] psig K,

=

[*]

K2

=

[*]

1(3=

[*]

"Ti

=

[*1 sec.

"T2

=

[*] sec.

f(AI)

=

An even function of the indicated difference between top and bottom detectors of the powerrange nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where q% and qb are the percent power in the top and bottom halves of the core respectively, and qt + qb is total core power in percent of RATED POWER, such that:

1. For qt - qb within [*,[*

,f (AI) = 0.

2.

For each percent that the magnitude of q, - qb exceeds [*I % the AT trip setpoint shall be automatically reduced by an equivalent of [*] %

of RATED POWER.

3.

For each percent that the magnitude of qt - qb exceed -[*] % the AT trip setpoint shall be automatically reduced by an equivalent of [*] %

of RATED POWER.

Note: r] As specified in the COLR LAR 187 TS 2.3-2 07T267=0

B.

Overpower AT<-ATo[ K 4 -K 5 r3S T-K 6(T-T')-f(AI)]

where AT0

= Indicated AT at RATED POWER, % RATED POWER T

= A^..ag.

TemperatureReference Average Temperature at RATED POWER, 'F T'

<=[*]OF K(4 K5

_ [*] for increasing T; [*] for decreasing T 16

_ [*]for T > T; [*] for T < T T3

= [*] sec.

f(AI)

= As defined aboO for all Ai Note: [M As specified in the COLR

4.

Reactor Coolant Flow A.

Low reactor coolant flow per loop _> 90% of normal indicated flow as measured by elbow taps.

B.

Reactor coolant pump motor breaker open

1.

Low frequency setpoint Ž_ 55.0 Hz

2.

Low voltage setpoint > 75% of normal voltage

5.

Steam Generators Low-low steam generator water level > 5% of narrow range instrument span.

LAR 187 TS 2.3-3 0712"727"2

c. Maximum Coolant Activity

1. The specific activity of the reactor coolant shall be limited to:

A. <*20-1.0 ICi/gram DOSE EQUIVALENT 1-131, and 91 /P*i B.

gross radioactivity due to nuclides with half-lives > 30 minutes E

cc excluding tritium ( E is the average sum of the beta and gamma energies in Mev per disintegration) whenever the reactor is critical or the average coolant temperature is > 5000F.

2. If the reactor is critical or the average temperature is > 500°F:

A. With the specific activity of the reactor coolant > 0.-201.0 pCi/gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or exceeding.

the limit *hW-n on Figur TTS 3.

3650 uCi/aram DOSE EQUIVALENT 1-131, be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature of < 500OF within six hours.

91 'Uci B. With the specific activity of the reactor coolant >---=-

of gross radioactivity, E

cc be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature < 500OF within six hours.

C. With the specific activity of the reactor coolant > 0,-2-01.0 pCi/gram DOSE EQUIVALENT 1-131 or > 9 C perform the sample and analysis requirements E

cc of Table TS 4.1-2, item 1.f, once every four hours until restored to within its limits.

3. Annual reporting requirements are identified in TS 6.9.a.2.D.

LAR 187 TS 3.1-6 07/26/2002

c. Condensate Storage Tank

1. The Reactor Coolant System shall not be heated > 350oF unless a minimum of 39,000 gallons of water is available in the condensate storage tanks.

2. If the Reactor Coolant System temperature is > 350oF and a minimum of 39,000 gallons of water is not available in the condensate storage tanks, reactor operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3. If the time limit of TS 3.4.c.2 above cannot be met, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate action to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve and maintain the Reactor Coolant System temperature < 350oF within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. Secondary Activity Limits

1. The Reactor Coolant System shall not be heated > 350oF unless the DOSE EQUIVALENT Iodine-131 activity on the secondary side of the steam generators is

< 0.1 ý.Ci/ecgram.

2. When the Reactor Coolant System temperature is > 350oF, the DOSE EQUIVALENT Iodine-131 activity on the secondary side of the steam generators may exceed 0.1 ý+/-Ci/ez_.ram for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3. If the requirement of TS 3.4.d.2 cannot be met, then within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve and maintain the Reactor Coolant System temperature < 350oF within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LAR 187 TS 3.4-3 07726=

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.

OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

SPECIFICATION

a.

Shutdown Reactivity When the reactor is subcritical prior to reactor startup, the SHUTDOWN MARGIN shall be at least that as specified in the COLR

b.

Power Distribution Limits

1. At all times, except during Low Power Physics Tests, the hot channel factors defined in the basis must meet the following limits:

A. FQN(Z) Limits shall be as specified in the COLR.

B. FaN Limits shall be as specified in the COLR.

2. If FIHN exceeds its limit:

A. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either, restore F HNto within its limit or reduce thermal power to less than 50% of RATED POWER anid reduce the Power Range Neutron Flux-High Trip Setpoint to _55% of RATED POWER within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

B. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the limit, verify through flux maDoing that FHN has been restored to within the limit, or reduce thermal power to < 5% of RATED POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and C. Identify and correct the cause of the out-of-limit condition prior to increasing thermal power above the reduced thermal power limit reguired by action A and/or B. above.

Subseauent power operation may proceed provided that F4N is demonstrated, through incore flux mapping, to be within its limit prior to exceding the following thermal power levels:

i.

A nominal 50% of RATED POWER, ii. A nominal 75% of RATED POWER, and iii. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining _95% of RATED POWER LAR 187 TS 3.10-1 07/26/2002

33. If FN(Z) exceeds its limit:

A. Reduce the thermal power at least 1% for each 1% F9N(z) exceeds its limit within 15 minutes after each determination and similarly reduce the Power Range Neutron Flux-High Trip Setpoints and the Overpower A T Trip Setooints within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by at least 1 % for each 1% FoN(Z) exceeds its limit.

B. Identify and correct the cause of the out-of--limit condition prior to increasing thermal power above the reduced thermal power limit required by action A. above. Thermal power may be increased provided FN(Z) is demonstrated. throuah incore flux mapping to be within its limit.

J L. L I.

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I

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n

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COL.R ar. not truo, then roacto. po*Wo.

h.All he rduc. d by a kfacAienal amount of the d.ign p.w..

to a

,aluo for which tho rol io..

aro s ye tru, and the high nAoLtMn flMu trip cotpo9it shall be roducoed by the samo fractional R.Mount& If cUb~eguont inoomapnannot, within aR 24 hour_1 poriod, domonc~trato Ath-tho ho channol factorc aro met, thon the ovorpowogr AT And oportomporatur AT trip cetpoints shall bo cimilarly drdud 3A4.

Following initial loading and at regular effective full-power monthly intervals thereafter, power distribution maps using the movable detection system shall be made to confirm that the hot channel factor limits of TS 3.10.b.1 are satisfied.

5. 4_The measured FQEa(Z) hot channel factors under equilibrium conditions shall satisfy the relationship for the central axial 80% of the core as specified in the COLR.

6. &.Power distribution maps using the movable detector system shall be made to confirm

.the relationship of FaEQ(Z) specified in the COLR according to the following schedules with allowances for a 25% grace period:

A. During the comparison of incore to axial flux difference or targot flu* difforonco dotorminati*n or once per effective full-power monthly interval, whichever occurs first.

B. Upon achieving equilibrium conditions after reaching a thermal power level > 10%

higher than the power level at which the last power distribution measurement was performed in accordance with TS 3.10.b.5.A.

C. If a power distribution map indicates an increase in peak pin power, FON, of 2% or more, due to exposure, when compared to the last power distribution map, then either of the following actions shall be taken:

iL FFEQ(Z) shall be increased by the penalty factor specified in the COLR for comparison to the relationship of FaEQ(Z) specified in the COLR, or ii.

FQEQ(Z) shall be measured by power distribution maps using the incore movable detector system at least once every seven effective full-power days until a power distribution map indicates that the peak pin power, FaN, is not increasing with exposure when compared to the last power distribution map.

LAR 187 TS 3.10-2 07/26/2002

J7. If, for a measured FQEQ, the relationships of FQEQ(Z) specified in the COLR are not satisfied and, the r,,laW,,nhp, of F

-d(Z-)a

.o..d

.i the COLR aro satisfied, then within 12 houre take eAs-ef.-the following actions:

A. Reduce the axial flux difference limit at least 1% for each 1% F EQ (Z) exceeds its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after each determination and reduce the Power Ranae Neutron Flux-High Tridp Setpoints and the Overpower A T TriD Setooints within the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by at least 1% for each 1 % that the maximum allowable power of the axial flux difference limits is reduced.

LAR 187 TS 3.10-3 07/26/2002

B. Confirm that the hot channel factor limits of TS 3.10.b.1 are satisfied prior to increasing thermal power above the maximim allowable power of the axial flux difference limits.

8. Axial Flux Difference NOTE:

The axial flux difference shall be considered outside limits when two or more operable excore channels indicate that axial flux difference is outside limits.

A. Durinq power operation with thermal power > 50 Dercent of RATED POWER. the axial flux difference shall be maintained within the limits specified in the COLR.

If the axial flux difference is not within limits, within 15 minutes restore to within limits. If this action and associated completion time is not met, reduce thermal power to less than 50% RATED POWER within 30 minutes and reduce the Power Ranae Neutron Flux-High Trd setgoints _<

55 percent of RATED POWER within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

B. If the alarms used to monitor the axial flux difference are rendered inoperable. verify that the axial flux difference is within limits for each operable excore channel once within one hour and every hour thereafter.

A.Tako corroctive actiene to improeY the poWor distribution and upon acshieving equilibrium condeitioFn macuro the target flux differonea and verify that tho rolationchips of-Fai'RP cpocifiod in the COLR aro satisfied, OR B.Roduco reactor poWor and tho high neut9ro flux trip cotpoint by 1 % for oaeh poron th-at the loft hand sidec, of the rolatiOnchipc, of FQq(Z) pocie6ffo n h CL oc tho limitc sepocfied in tho right hand sides. Reactor powor may cUbcoquontly be icood provided that a powor dictrib9ution m~ap Yerifioc that tho rolationchipc of E4(Z) 6pecifiod in the COL=R ar saticfied with at loact 1% of mnargin for each peroont of power-level to e 18fincroaced.

7--Tho roforoncoe equilibrium indicated axial flux diffoenca as a function of poere level (called the target flux differenco) shall be moeasured at least onco por fuill power month:.

84The indicated axial flux differen shall 198 concidorod outsido of the limitc of T-9 3.10G.b.9 through TS 3.40.b.42 when moree than oeR of the OPERABLE oxcoroe ehannolc are, inRdicating the axial flux diffeFr~en to be eutsido a limit.

Q.&expt during physicc tests, during emeero detector calibration and excapt as moedified by T-9 3.49.b.40 through TS 3.49.b.12, the indicated axial flux di#fforen chall be main.taineAd-wit-hin the t-arget band about the target flux difforonco as Gpeeiliod i h

10. At a power level 00%, of rated PGS, ith neaed axial flux diffeFrene doviate6 15 minGutec or reactor peF cal b eded to a l8vel no greater than 00% of rated peweF.

LAR 187 07/26/2002 TS 3.10-4

Itpwrleoelc > 50% and 5 00% of ratod r

A. Tho iner-diAtAd ax-i;al fu diAforn*co may doviato from tho targot band, cpocifiod iR tho COL=R, for a m~axfiF.mum o-f ono ho.ur (cumulativo) in any 21 hour2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months podoed providodth flux diffo~Rono dooc6 not oxAAcoo tho outor onvolopo cpodfiod intho CQLR. If tho cumulatiKoe oxcoodc ono hour, then tho ator hall b38oeducod to m*50% of rated theFrmal pOwer8 within 30 mninutoc and tho high noutero flux cotpoint e-.*.-ued te -

55% of rated power.

If the-indicated axial flux diffo~ronc oxcooedc tho eutor onvelopo cpoeified in tho COL=R, thon tho roac~tor powor chall bo roduc-Ad to *50'6%

of ratod thermnal power wit.hin 30 minutoc, an th hghnutron flux cotpeint roducod to,-!W

56%ef-ate peweF.

B.A Powo"r iro.eas to a lovel > 00% of rated poWor ic con*tin.gont upo n he anesated ax-ial,_ nux ainurun o n u.*

,l*

.....iunn, *,**-.'**

1-2.

At a; owor 1oYo1 noG aroator man 54%

of rata I9owo A.T-ho indicatod axial flux 9oronco m~ay aoiato rrom 1W targot Dana.

B.A powor inRoFaco to a lovol >50% of ratod powor ic contin~gont upon tho indicatod axial flux diffaronco nlot boing oulcido itc targot band for moroF than Mex hourc (cumulatieo) of tho p*rocding 24 21 horpoiod.

OnR half of tho timo th*

indicatod axial flux difforonco8 icS Ut of itc targoet band, Up to 50% of Fatod poweo ri to bo scuntod ac c*ntributig* to th* ono hour cumulativo maiu tho flux di#FfforoFncmay doviato fromn itc targoet band at a powor lovol 13.AlaFrmc chall normFFally roquiFBromot Of T-9 3.2 If tho alarm aro tor leggod, and cl*Wnfola half hourly thoroaftor bo8 ucod to indicato no l0.b.1 or 9 tho flux dfifforor nporaily out of crelic, nconformnanco with the flux diffoencoI~

co timo requiromon~t of T-S 3. 10. b. 1 !.A.

thon tho axial flux difforoncl chall bo od, oor; hour for the firet 21 ho*ur, and

c.

Quadrant Power Tilt Limits

1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:

A.

Eliminate the tilt.

B.

Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then reduce power to 50% or lower.

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:

LAR 187 07/26/2002 TS 3.10-5 BWBF' A P99 lW "Mli Me l

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Quadrant Power Tilt Monitor If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the quadrant tilt shall be logged once per shift and after a load change > 10% of rated power or after > 24 steps of control rod motion. The monitors shall be set to alarm at 2% tilt ratio.

k.

Core Average Temperature During steady-state power operation, T, shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n.

Reactor Coolant System Pressure During steady-state power operation, Reactor Coolant System pressure shall be maintained within the limits specified in the COLR, except as provided by TS 3.1O.n.

m.

Reactor Coolant Flow

1. During steady-state power operation, reactor coolant total flow rate shall be >

O3,000178,000 gallons per minute average pe.,leep and greater than or equal to the limit specified in the COLR. If reactor coolant flow rate is not within the limits as specified in the COLR, action shall be taken in accordance with TS 3.1O.n.

2. Compliance with this flow requirement shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING, betW::o.

70%at or above 90% and 95% power with plant parameters as constant as practical.

n.

DNBR Parameters If, during power operation any of the conditions of TS 3.10.k, TS 3.10.1, or TS 3.10.m.1 are not met, restore the parameter in two hours or less to within limits or reduce power to < 5% of thermal rated power within an additional six hours. Following analysis, thermal power may be raised not to exceed a power level analyzed to maintain a DNBR greater than the minimum DNBR limit.

LAR 187 TS 3.10-9 07/26/2002

TS FIGURE TS 3.1-3 DELETED LAR 187 07/26/2002 Page 1e4

5.2 CONTAINMENT APPLICABILITY Applies to those design features of the Containment System relating to operational and public safety.

OBJECTIVE To define the significant design features of the Containment System.

SPECIFICATION

a. Containment System

1. The Containment System completely encloses the entire reactor and the Reactor Coolant System and ensures that leakage of activity is limited, filtered and delayed such that off-site doses resulting from the design basis accident are within the guidelines of 10 CFR Part 40050.67. The Containment System provides biological shielding for both normal OPER'TING conditions and accident situations.

2. The Containment System consists of:

A. A free-standing steel reactor containment vessel designed for the peak pressure of the design basis accident.

B. A concrete shield building which surrounds the containment vessel, providing a shield building annulus between the two structures.

C. A Shield Building Ventilation System that causes leakage from the reactor containment vessel to be delayed and filtered before its release to the environment.

D. An Auxiliary Building Special Ventilation System that serves the special ventilation zone and supplements the Shield Building Ventilation System during an accident condition by causing any leakage from the Residual Heat Removal System (RHRS) and certain small amounts of leakage that might be postulated to bypass the Shield Building Ventilation System to be filtered before their release.

LAR 187 TS 5.2-1 07726=2002

5.3 REACTOR CORE APPLICABILITY Applies to the reactor core.

OBJECTIVE To define those design features which are essential in providing for safe reactor core operations.

SPECIFICATION

a. Fuel Assemblies The reactor shall contain 121 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or ZIRLOQT clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material.

Limited substitutions of zirconium alloy, ZIRLOTM, or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead-test-assemblies that have not completed representative testing may be placed in non-limiting core regions. Lead-test-assemblies shall be of designs approved by the NRC for use in pressurized water reactors and their clad materials shall be the materials approved as part of those designs.

b. Control Rod Assemblies The reactor core shall contain 29 control rod assemblies. The control material shall be silver indium cadmium.

LAR 187 TS 5.3-1 07/26/2002

(3)

Nissley, M.E. et, al., "Westinghouse Large-Break LOCA Best Estimate Methodology," WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, March 1991, Volume 1:

Model Description and Validation; Addendum 4: Model Revisions.

(4)

N. Lee et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-1 0081-NP-A (Non-Proprietary), dated August 1985.

(5)

C.M. Thompson, et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condesation Model,"

WCAP-10054-P-A, Addendum 2, Revision 1 (Proprietary) and WCAP-10081-NP (Non-Proprietary), dated July 1997.

(6)

XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, dated October 1986.

(7)

ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Meghodology for Rod Burnups of 62 GWd/MTU,"

Advanced Nuclear Fuels Corporation, dated December 1991.

(8)

EMF-92-116 (P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, dated February 1999.

(9)

XN-NF-77-57, Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II, dated January 1978, and Supplement 2, dated October 1981.

(10)

WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," February 1994 (W Proprietary).

(11)

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATIO MEHDOLGY-,"July185 (W Proprietary)

(12)

WCAP-8745-P-A, Design Bases for the Thermal Overtemperature AT and Thermal Overpower AT trip functions, September 1986.

(13)

WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES-TOPICAL REPORT,"

September 1974 (Westinghouse Proprietary).

(14)

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995 (Westinghouse Proprietary).

LAR 187 TS 6.9-4 7726120=2

BASIS - Safety Limits-Reactor Core (TS 2.1)

To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all OPERATING conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters of RATED POWER, reactor coolant temperature and pressure have been related to DNB through a DNB correlation. The DNB correlation has been developed to predict the DNB heat flux and the location of the DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to the DNBR limit. This minimum DNBR corresponds to a 95%

probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all OPERATING conditions.

The SAFETY LIMIT curves as provided in the Core Operating Report Limits Report which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation) represent the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNBR is equal to the DNBR limit or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNBR ratio reaches the DNBR limit and thus, this limit is conservative with respect to maintaining clad integrity. The area where clad integrity is ensured is below these lines.

The curves are based on the nuclear hot channel factor limits of as specified in the COLR.

These limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion.

The control rod insertion limits are given in TS 3.10.d. Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits as specified in the COLR ensure that the DNBR is always greater at partial power than at full power.

The Reactor Control and PROTECTION SYSTEM is designed to prevent any anticipated combination of transient conditions that would result in a DNBR less than the DNBR limit.

Twe-Three departure from nucleate boiling ratio (DNBR) correlations used in the safety analyses:

the WRB-1 DNBR correlation, the high thermal performance (HTP) DNBR correlation and the W-3 DNBR correlation. The WRB-1 correlation applies to the Westinghouse 422 V+ fuel. The HTP correlation applies to FRA-ANP fuel with HTP spacers. The W-3 correlation is used when the coolant conditions are outside the range of the WRB-1 correlation or for the analysis of non-HTP FRA-ANP fuel designs and for all fuel designs at low pressure and temperature conditions (e.g., the conditions analyzed during a main steam line break accident). Beth-_DNBR correlations have been qualified and approved for application to Kewaunee. The DNBRmI,-m'I'A limits are 1.14 for the HTP correlation, 1.17 for the WRB-1 correlation, and 1.30 for the W-3 correlation.

LAR 187 TS B2.1-1 07/26/202

BASIS - Safety Limit - Reactor Coolant System Pressure (TS 2.2)

The Reactor Coolant System(1 ) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the Reactor Coolant System is the primary barrier against the release of fission products. By establishing a system pressure limit, the continued integrity of the Reactor Coolant System is ensured. The maximum transient pressure allowable in the reactor pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the Reactor Coolant System piping, valves and fittings under USASI B.31.1.0 is 120% of design pressure. Thus, the SAFETY LIMIT of 2735 psig (110% of design pressure, 2485 psig) has been established. (2)

The settings of the power-operated relief valves, the reactor high pressure trip and the safety valves have been established to prevent exceeding the SAFETY LIMIT of 2735 psig. for all transients except the hypothetical RCCA Ejection accident, for which the faulted condition stre-ss limit acceptance criterion of 3105 psig (3120 psia) is applied.

The initial hydrostatic test was conducted at 3107 psig to ensure the integrity of the Reactor Coolant System.

(1) USAR Section 4 (2) USAR Section 4.3 LAR 187 07726/200 TS B2.2-1

Maximum Coolant Activity (TS 3.1.c)

The maximum dose to tho thyroid and whoel body that an individual may receive following an accident is specified in GDC 19 and 10 CFR=-0050.67. The limits on maximum coolant activity ensure that the calculated doses are held to the limits specified in GDC 19 and to a fraction of the 10 CFR 40050.67 limits.

The Reactor Coolant Specific Activity is limited to < 0421.0 I.Ci/gram DOSE EQUIVALENT 1-131 to ensure the thy..iddose does not exceed the GDC-19 and 10 CFR 40050.67 guidelines. The applicable accidents identified in the USAR('5) are analyzed assuming an RCS activity of 0"1.0 RCi/gram DOSE EQUIVALENT 1-131 incorporating an accident initiated iodine spike when required.

-To ensure the conditions allowed by Fig..o TS. 3.1 3are taken into account, the applicable accidents are also analyzed considering a pre-existing iodine spike of 60 i, Ci/.ram DOSE EQUIVALENT 1-131.

b,,Aod, on, Fig...

T-8 3.4 3.

The results obtained from these analyses indicate that the control room and off-site thyie..,doses are within the acceptance criteria of GDC 19 and a fraction of 10 CFR 40050.67 limits.

9 1 'iCi The Reactor Coolant Specific Activity is also limited to a gross activity of _< 9-=_

Again the E cc 91 XCi accidents under consideration are analyzed assuming a gross activity of 9_

. The results E

cc obtained from these analyses indicate the control room and off-site whole bedy dose are within the acceptance criteria of GDC-19 and a small fraction of 10 CFR 40050.67 limits.

The action of reducing average reactor coolant temperature to < 500OF prevents the release of activity should a steam generator tube rupture occur since the saturation pressure of the reactor coolant is below the lift pressure of the main steam safety valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

(15) USAR Section 14.0 LAR 187 TS B3.1-8 07/26/2002

BASIS - Steam and Power Conversion System (TS 3.4)

Main Steam Safety Valves (TS 3.4.a)

The ten main steam safety valves (MSSVs) (five per steam generator) have a total combined rated capability of 7,660,380 lbs./hr. at 1181 Ibs./in.2 pressurc. Tho maximum full p'Wor steam flow at 1721 1780 MWt i* 7,449,0007L7Z0 00 lbc./hr. This flow ensures that the main steam pressure does not exceed 110 percent of the steam generator shell-side design pressure (the maximum pressure allowed by ASME B&PV Code) for the worst-case loss-of-heat-sink event.

Thoroforo, tho-main comaft al'.~c will be _habl orlot ttoa maImu ctaM flow i While the plant is in the HOT SHUTDOWN condition, at least two main steam safety valves per steam generator are required to be available to provide sufficient relief capacity to protect the system.

The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Plan.

Auxiliary Feedwater System (TS 3.4.b)

The Auxiliary Feedwater (AFW) System is designed to remove decay heat during plant startups, plant shutdowns, and under accident conditions. During plant startups and shutdowns the system is used in the transition between Residual Heat Removal (RHR) System decay heat removal and Main Feedwater System operation.

The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow from the AFW pumps to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE, each capable of taking suction from the Service Water System and supplying AFW to separate steam generators.

The turbine-driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the main steam isolation valves and shall be capable of taking suction from the Service Water System and supplying AFW to both of the steam generators. With no AFW trains OPERABLE, immediate action shall be taken to restore a train.

Auxiliary feedwater trains are defined as follows:

"A" train -

"A" motor-driven auxiliary feedwater pump and associated AFW valves and piping to "A" steam generator, not including AFW-1 OA or AFW-1 OB "B" train -

"B" motor-driven auxiliary feedwater pump and associated AFW valves and piping to "B" steam generator, not including AFW-1OA or AFW-1OB Turbine-driven Turbine-driven AFW pump and associated AFW valves and piping to both train -

"A" steam generator and "B" steam generator, including AFW-1OA and AFW-10B LAR 187 TS B3.4-1 07/26/2002

In the unlikely event of a loss of off-site electrical power to the plant, continued capability of decay heat removal would be ensured by the availability of either the steam-driven AFW pump or one of the two motor-driven AFW pumps, and by steam discharge to the atmosphere through the main steam safety valves. Each motor-driven pump and turbine-driven AFW pump is normally aligned to both steam generators. Valves AFW-10A and AFW-10B are normally open. Any single AFW pump can supply sufficient feedwater for removal of decay heat from the reactor.

As the plant is cooled down, heated up, or operated in a low power condition, AFW flow will have to be adjusted to maintain an adequate water inventory in the steam generators. This can be accomplished by any one of the following:

1.

Throttling the discharge valves on the motor-driven AFW pumps

2.

Closing one or both of the cross-connect flow valves

3.

Stopping the pumps If the main feedwater pumps are not in operation at the time, valves AFW-2A and AFW-2B must be throttled or the control switches for the AFW pumps located in the control room will have to be placed in the "pull out" position to prevent their continued operation and overfill of the steam generators. The cross-connect flow valves may be closed to specifically direct AFW flow. Manual action to re-initiate flow after it has been isolated is considered acceptable based on an evaluationaealysee performed by A.PSC-C and the Westinghouse Electric Company.

//C~~per..tinn. This evaluationThesoc analy.........

n ativly a..umod the plant...a. At 100%

Fi"i,.P.e-- and demonstrated that operators have at least 10 minutes to manually initiate AFW during any design basis accident below 15% of RATED POWER with no steam generator dryout.

or= Gem damaipreactor coolant system overpressure. The placing of the AFW control switches in the "pull out" position, the closing of one or both cross-connect valves, and the closing or throttling of valves AFW-2A and AFW-2B are limited to situations when reactor power is <15% of RATED POWER to pr. ido furthor* marg.in in the analycis.

During accident conditions, the AFW System provides three functions:

1.

Prevents thermal cycling of the steam generator tubesheet upon loss of the main feedwater pump

2.

Removes residual heat from the Reactor Coolant System until the temperature drops below 300-350°F and the RHR System is capable of providing the necessary heat sink

3.

Maintains a head of water in the steam generator following a loss-of-coolant accident Each AFW pump provides 100% of the required capacity to the steam generators as assumed in the accident analyses to fulfill the above functions. Since the AFW System is a safety features system, the backup pump is provided. This redundant motor-driven capability is also supplemented by the turbine-driven pump.

The pumps are capable of automatic starting and can deliver full AFW flow within one minute after the signal for pump actuation. Howovor, a.alyc..f rm full poWor dom,.'nctrat that initiation of fow can be delayed for at oaIct 10 m.nutoc With no cstam g"n.rato. dro'ut or.. ro damago. The head generated by the AFW pumps is sufficient to ensure that feedwater can be pumped into the steam generators when the safety valves are discharging and the supply source is at its lowest head.

LAR 187 TS B3.4-2 07/26/2002

Analyses by WPSC and the Wo.tinghoo*

u El.. tr Ceorpration

,.ho.

that AFW

,2A.

and. AW-^ 2B may be in tho throttlo~d or clocod pocition, or the AF-W pump control AAwitchoc lecatod in tho control roomA may bo in the "pull out"' pocition MAthout aRoprm to saft*,. Thicr dooc not concMtitt -a condi6tion of inprblt lco nT-8 3.4.b. 1 or TS 2-4.b.2. T-ho anaflycic shows that di.orcos automatic roactor trp6 oncuro a plant trip boforo any coro damage or cyctem ov"rproccUro occur and that at lo-ast 10Q Min'utoc aro ayallabl for. tho opoatorc to manually initiato au*iliar1' foodwator flow (sta~t AR!

pumApc or fully opon A.R.A. 2-A. _and APXA 2B) for any G~edible aseidentfrma ntl to meet single failure criteria. In a secondary line break, it is assumed that the pump discharging to the intact steam generator fails and that the flow from the redundant motor-driven AFW pump is discharging out the break. Therefore, to meet single failure criteria, the turbine-driven AFW pump was added to Technical Specifications.

The cross-connect valves (AFW-10A and AFW-10B) are normally maintained in the open position This provides an added degree of redundancy above what is required for all accidents except for a MSLB. During a MSLB, one of the cross-connect valves will have to be repositioned regardless if the valves are normally opened or closed. Therefore, the position of the cross-connect valves does not affect the performance of the turbine-driven AFW train. However, performance of the train is dependent on the ability of the valves to reposition. A,,hetigh-analyses have demonstrated that operation with the cross-connect valves closed is acceptable when reacctor power is-the--T r*-ctric op..atio, with the

'alve., l... d to <15% of RATED POWER. At > 15% RATED POWER, closure of the cross-connect valves renders the TDAFW train inoperable.

An AFW train is defined as the AFW system piping, valves and pumps directly associated with providing AFW from the AFW pumps to the steam generators. The action with three trains inoperable is to maintain the plant in an OPERATING condition in which the AFW System is not needed for heat removal. When one train is restored, then the LIMITING CONDITIONS FOR OPERATION specified in TS 3.4.b.2 are applied. Should the plant shutdown be initiated with no AFW trains available, there would be no feedwater to the steam generators to cool the plant to 350°F when the RHR System could be placed into operation.

It is acceptable to exceed 350°F with an inoperable turbine-driven AFW train.

However, OPERABILITY of the train must be demonstrated within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after exceeding 350°F or a plant shutdown must be initiated.

Condensate Storage Tank (TS 3.4.c)

The specified minimum water supply in the condensate storage tanks (CST) is sufficient for four hours of decay heat removal. The four hours are based on the Kewaunee site specific station blackout (loss of all AC power) coping duration requirement.

The shutdown sequence of TS 3.4.c.3 allows for a safe and orderly shutdown of the reactor plant if the specified limits cannot be met. (1)

(1) USAR Section 8.2.4 LAR 187 TS B3.4-3 07/26/2002

Secondary Activity Limits (TS 3.4.d)

The maximum dose to the thyroid and wholo body that an individual may receive following an accident is specified in GDC 19 and 10 CFR400 50.67. The limits on secondary coolant activity ensure that the calculated doses are held to the limits specified in GDC 19 and to a fraction of the 10 CFR 40050.67 limits.

The secondary side of the steam generator's activity is limited to < 0.1

.iCi/eeram DOSE EQUIVALENT 1-131 to ensure the thyFeid dose does not exceed the GDC-19 and 10 CFR 4-00 50.67 guidelines. The applicable accidents identified in the USAR(2) are analyzed assuming various inputs including steam generator activity of 0.1 pCi/ee-aram DOSE EQUIVALENT 1-131.

The results obtained from these analyses indicate that the control room and off-site thyei

.de"e-doses are within the acceptance criteria of GDC-1 9 and a fraction of 10 CFR 400-50.67 limits.

(2) USAR Section 14.0 LAR 187 07/26/2002 TS B3.4-4

FQN(Z), Height Dependent Nuclear Flux Hot Channel Factor FQN(Z), Height Dependent Nuclear Flux Hot Channel Factor, is defined as the maximum local linear power density in the core at core elevation Z divided by the core average linear power density, assuming nominal fuel rod dimensions.

FQEQ(Z) is the measured FQN(Z) obtained at equilibrium conditions during the target flux determination.

An upper bound envelope for FQN(Z) as specified in the COLR has been determined from extensive analyses considering all OPERATING maneuvers consistent with the Technical Specifications on power distribution control as given in TS 3.10. The results of the loss-of-coolant accident analyses based on this upper bound envelope indicate the peak clad temperatures_, with a high probability.

remain less than the 22000 F limit.

The FQN(Z) limits as specified in the COLR are derived from the LOCA analyses. The LOCA analyses are performed for Westinghouse 422 V-+ fuel FRA-ANP heavy fuel and for FRA-ANP standard fuel.

When a FN(Z) measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

FQN(Z) is arbitrarily limited for P < 0.5 (except for low power physics tests).

FH Nuclear Enthalpy Rise Hot Channel Factor FHN, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the maximum integral of linear power along a fuel rod to the core average integral fuel rod power.

It should be noted that FAHN is based on an integral and is used as such in DNBR calculations.

Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal power shape at the point of maximum heat flux is not necessarily directly related to FaN.

The F N limit is determined from safety analyses of the limiting DNBR transient events. The safety analyses are performed for FRA-ANP heavy fuel., aRd feF FRA-ANP standard fuel. and Westinghouse 422 V-+ fuel. In these analyses, the important operational parameters are selected to minimize DNBR. The results of the safety analyses must demonstrate that minimum DNBR is less than the DNBR limit for a fuel rod operating at the F N limit.

The use of F N in TS 3.1O.b.5.C is to monitor "upbum" which is defined as an increase in FN with exposure. Since this is not to be confused with observed changes in peak power resulting from such phenomena as xenon redistribution, control rod movement, power level changes, or changes in the number of instrumented thimbles recorded, an allowance of 2% is used to account for such changes.

LAR 187 TS B3.10-2 07/26/2002

Rod Bow Effects No penalty for rod bow effects need be included in TS 3.10.b.1 for FRA-ANP fuel.(')

Surveillance Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identifies operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met.

These conditions are as follows:

1.

Control rods in a single bank move together with no individual rod insertion differing by more than an indicated 12 steps from the bank demand position where reactor power is Ž 85%, or an indicated 24 steps when reactor power is < 85%.

2.

Control rod banks are sequenced with overlapping banks as specified in the COLR.

3.

The control bank insertion limits as specified in the COLR are not violated, except as allowed by TS 3.1O.d.2.

4.

4--The axial power distribution, expressed in terms of axial flux difference, is maintained within the limits.

The limits on axial flux difference (AFD) assure that the axial power distribution is maintained such that the FQ(Z) upper bound envelope of FQLIMIT times the normalized axial peaking factor [K(Z) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. This ensures that the power distributions assumed in the large and small break LOCA analyses will bound those that occur during plant operation.

Provisions foref monitoring the AFD on an automatic basis are derived from the glant process computer through the AFD monitor grogram. The computer determines the AFD for each of the operable excore channels and provides a computer alarm if the AFD for at least 2 of 4 or 2 of 3 operable excore channels are outside the AFD limits and the reactor Dower is greater than 50 percent of RATED POWER.Ax...a..

d

,m.tibution conro cp. ifiatoc Which are given in t-,m of flu* diffoFronc control and conrol1 bank incortion "imitS arebcorvod. Flux diffo~renc roforc to the dieffooc in igns between the top and bottom halves of two soction oxcoro noutront dotoctore. The flux difrna ica eacure of the axial offsot whicsh is defined as the diffo~ronc in normalize powor between tho top and bottom halyoc Of the core.

('N E. Hoppe, "Mechanical Design Report Supplement for Kewaunee High Bumup (49 GWD/MTU)

Fuel Assemblies," XN-NF-84-28(P), Exxon Nuclear Company, July 1984.

LAR 187 TS B3.10-3 07/26/2002

coo 0

CN C14 r

cli c c C

c

Tho targot (Or reforenco) valup8 of fluxW dfiffronco is dotorm~inod_ ;at follows. At any time that oguilibrum xenon conditionc, havo boon ostablished, tho Rindiatod flu diforoccdtrio fAro tho nuc-loar intmontatien. TPhic Value, divided by tho kifratin of full power at which the coro wacO (PERATI NG i tho full powor value of the tagtflxdfeeronc.

VR-alues for al torcr power 6 Woo e

r obtained by mnultiplying tho full pwrvuoby tho fractioal powor. Sincoe tho indicatod oguilibrium value' iwat notod, no alWManos for excoro detector orror are necossary and indicsatod devaAtiont RAtseiedn tho COL=R aro permitted fromA the indicatod rofoensoe value.

Strict con-troMl o-f tho flUx d~iffoAronso (and rod poskitin) it not atnccar,'Rrss~

duFrig part power operation. This it bocause xonon dittributio control at part poere it not as significntRM ats the control at full powor and allowance hat boon mnade in predicting the hoat flux peaking factorsfo losst strict control at Part powor. Stric~t control of tho flux diffreronc it not pocciblo during cortainf physiGc, tts or duri;ng re*i, poriodic, excoro cal;i*batio*n which roi r; I.e larger flux diffelRoos than permi~tte-d. Th~erefoe tho seiiations on powor distribution conrol1 aro not applied during physics tests o9r oxcoe

rca*atioFn; this it acceptable duo to tho loW fprbability of a signifieRat

_;accidont occurring during those oporations.

In somo instancos of rapid plant power roduction automatic rod motion will cauco the flux di~ffonc to deviato fro~m tho target ban~d whon tho roducod poworA l9vo0A it r~achod. Thit dootno nocetarly ffos th xoon ictribution suffieiontly to chango the envelope of poaking factort which can boA FAroasho-d on a subto6quont rotum to full poWe orithin the tar got band; howoyor, to iApli~f,' the specification, a limfitation of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in an peid of 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months is placod on; oporation outGido tho band. Thit onturos that tho rosulting xonon81 d-isribtions49 are not significantly diffo~rot from thosoresul~ting fromA oporation within tho targlet band.

Tho intatnou cnequoncots of boing outsido tho band, provided rod intortion liMit arc observod, it not worco than a 10% inro~mont in poaking factor for flux difo~;ronc onvelope at specified in tho COLR. Thorofor, while tho deviation oxistts the power level is limited to 00% or lower depending en the indicrated_ flu-x difference without additional core monitoring. if, for any roaten, flux differenee is net controlled within the target ban~d as specified in; the COLR frasln a period as one hour, then xenon distr~ibuden may be significantly changed and operaina 0 is required to protect against potentially morge saevee consequ~enes of som~e accidents unless As disc.usrsed above, the ess9ene of the precodure is to mnaintain the xenon distribution in the coee as close to the equilibrium full9W8 poerodition as possible. This is accomplished by using the For Condition 11 events the core is protested from overpower and a minimum DNBR less than the DNBR limit by an automatic Protection System. Compliance with the specification is assumed as a precondition for Condition 11 transients; however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.

LAR 187 07/26/2002 TS B3.10-5

Inoperable Rod Position Indicator Channels (TS 3.10.f)

The rod position indicator channel is sufficiently accurate to detect a rod +/- 12 steps away from its demand position. If the rod position indicator channel is not OPERABLE, then the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or movable incore detectors, will be used to verify power distribution symmetry.

Inoperable Rod Limitations (TS 3.10.q)

One inoperable control rod is acceptable provided the potential consequences of accidents are not worse than the cases analyzed in the safety analysis report. A 30-day period is provided for the reanalysis of all accidents sensitive to the changed initial condition.

Rod Drop Time (TS 3.10.h)

The required drop time to dashpot entry is consistent with safety analysis.

Core Average Temperature (TS 3.10.kA The RCS core average temperature limit is consistent with full power operation within the nominal operational envelope. Either Tavq control board indicator readinqgs or computer indications are averaged to obtain the value for comparison to the limit. The limit is based on the average of either 4 control board indicator readings or 4 computer indications. A higher Tava will cause the reactor core to approach DNB limits.

Reactor Coolant System Pressure (TS 3.10.m)

The RCS oressure limit is consistent with operation within the nominal operational envelope. Either pressurizer pressure control board indicator readings or comDuter indications are averaged to obtain the value for comparison to the limit. The limit is based on the average of either 4 control board indicator readings or 4 computer indications. A lower pressure will cause the reactor core to aporoach DNB limits.

Reactor Coolant Flow (TS 3.10.m)

The reactor coolant system (RCS) flow limit, as specified in the COLR, is consistent with the minimum RCS flow limit assumed in the safety analysis adiusted by the measurement uncertainty.

The safety analysis assumes initial conditions for plant parameters within the normal steady state envelope. The limits placed on the RCS pressure, temperature, and flow ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the analyzed tansients.

The RCS flow normally remains constant during an operational fuel cycle with all reactor coolant pumps running. At least two plant computer readouts from the loop RCS flow instrument channels are averaged Per reactor coolant loop and the sum of the reactor coolant loop flows are compared to the limit. Operating within this limit will result in meeting the DNBR criterion in the event of a DNB-limited event.

LAR 187 TS B3.10-8 07/26/2002

ATTACHMENT 3 Letter from M. E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 187 Revised Pages for Technical Specifications including Bases TS ii TS vi TS 1.0-6 TS 2.1-1 TS 2.3-2 through TS 2.3-3 TS 3.1-6 TS 3.4-3 TS 3.10-1 through TS 3.10-3 TS 3.10-6 TS Figure 3.1-3 TS 5.2-1 TS 5.3-1 TS 6.9-4 TS B2.1-1 TS B2.2-1 TS B3.1-1 TS B3.1-8 TS B3.4-1 through TS B3.4-4 TS B3.10-2 TS B3.10-3 TS B3.10-6 through TS B3.10-7

Section Title 3.2 Chemical and Volume Control System..........................................................

3.2-1 3.3 Engineered Safety Features and Auxiliary Systems......................................

3.3-1 3.3.a Accumulators...........................................................................

3.3-1 3.3.b Emergency Core Cooling System.............................................

3.3-2 3.3.c Containment Cooling Systems..................................................

3.3-4 3.3.d Component Cooling System.....................................................

3.3-6 3.3.e Service W ater System..............................................................

3.3-7 3.4 Steam and Power Conversion System........................................................

3.4-1 3.4.a Main Steam Safety Valves........................................................

3.4-1 3.4.b Auxiliary Feedwater System.....................................................

3.4-2 3.4.c Condensate Storage Tank........................................................

3.4-3 3.4.d Secondary Activity Limits..........................................................

3.4-3 3.5 Instrumentation System.............................................................................

3.5-1 3.6 Containment System.................................................................................

3.6-1 3.7 Auxiliary Electrical Systems........................................................................

3.7-1 3.8 Refueling Operations.................................................................................

3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits................................................

3.10-1 3.10.a Shutdown Reactivity...............................................................

3.10-1 3.10.b Power Distribution Limits........................................................

3.10-1 3.10.c Quadrant Power Tilt Limits......................................................

3.10-3 3.1O.d Rod Insertion Limits................................................................

3.10-4 3.10.e Rod Misalignment Limitations.................................................

3.10-5 3.10.f Inoperable Rod Position Indicator Channels........................... 3.10-5 3.10.g Inoperable Rod Limitations.....................................................

3.10-6 3.10.h Rod Drop Time.......................................................................

3.10-6 3.10.i Rod Position Deviation Monitor...............................................

3.10-6 3.10.j Quadrant Power Tilt Monitor...................................................

3.10-6 3.10.k Core Average Temperature....................................................

3.10-6 3.10.1 Reactor Coolant System Pressure..........................................

3.10-6 3.10.m Reactor Coolant Flow............................................................

3.10-6 3.10.n DNBR Parameters.:-.............................

3.10-6 3.11 Core Surveillance Instrumentation............................................................