Text

License Amendment Request Regarding Adoption of Relaxed Axial Offset Control (RAOC)
	ML051300330
Person / Time
Site:	Ginna
Issue date:	04/29/2005
From:	Korsnick M; Constellation Energy Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML051300330 (123)
	v • d • e

Maria Korsnick 1503 Lake Road Vice President Ontario, New York 14519-9364 585.771.3494 585.771.3943 Fax maria.korsnickaconstellation.com Constellation Energy R.E. Ginna Nuclear Power Plant April 29, 2005 Ms. Donna M. Skay Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request Regarding Adoption of Relaxed Axial Offset Control (RAOC)

R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Ms. Skay:

In accordance with the provisions of 10 CFR 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) is submitting a request for a license amendment to modify the Technical Specifications (TS) for the R.E. Ginna Nuclear Power Plant.

The proposed amendment would revise the TS to allow the use of the Relaxed Axial Offset Control (RAOC) methodology for certain Power Distribution Limits. The use of the RAOC methodology permits reduced operator actions to maintain compliance with power distribution control TS. The proposed changes to use RAOC are being requested to accommodate the planned power uprate. These changes are consistent with applicable RAOC requirements specified in Revision 3 to NUREG-1431, Standard Technical Specifications Westinghouse Plants, with the exception of one surveillance in LCO 3.2.4, Quadrant Powver Tilt Ratio, which will be maintained consistent with the current TS.

It has been determined that this amendment application does not involve a significant hazard consideration as determined by 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

Enclosure I provides a description and assessment of the proposed changes. Enclosure 2 provides the existing TS pages marked up to show the proposed changes. Enclosure 3 provides revised (clean) TS pages. Enclosure 4 provides the existing TS Bases pages marked up to reflect zoo 130i

the proposed changes (for information only). Changes to the TS Bases will be provided in a future update in accordance with the Bases Control Program. Enclosure 5 provides a list of regulatory commitments associated with this license amendment request.

Approval of this amendment application is requested by April 30, 2006 to provide adequate time to prepare for implementation. Once approved, the amendment will be implemented prior to startup from the fall 2006 refueling outage.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated New York State Official.

If you have any questions regarding this submittal, please contact George Wrobel, Nuclear Safety and Licensing at (585) 771-3535.

t I1 t

G.

M orsnick

Enclosures:

1. Evaluation of Proposed Change

2. Proposed Technical Specification Changes (markup)

3. Revised Technical Specification Pages (retyped)

4. Marked-up Copy of Technical Specification Bases

5. List of Regulatory Commitments

STATE OF NEWV YORK TO WIT:

COUNTY OF WAYNE I, Mary G. Korsnick, being duly swoom, state that I am Vice President - R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this response on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

TYWa4 2&1 swcL Subscribed and sworn before me, a Notary Public in and for the State of New York and County of B2t' e- ,this 42q dayof Ag ~ ,2005.

'WITNESS my Hand and Notarial Seal: %47x adzv 6 Notary Public My Commission Expires:

ii- DMICHALENE

/.. 1/ A BUNTS Date Notary Public, State of NewYork Registration No. 01 BU6018576 Monroe County ,

Commission Expires Jan IIJ 7O

Cc: Ms. Donna M. Skay (Mail Stop 0-8-C2)

Project Directorate I Division of Licensing Project Management Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Mr. Peter R. Smith New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223

ENCLOSURE I R.E. Ginna Nuclear Power Plant Description and Assessment of Proposed Change

Subject:

Revision to Technical Specifications 3.2.1, 3.2.3, 3.2.4, 3.3.1 and 5.6.5 to Adopt Relaxed Axial Offset Control Methodologies for Powver Distribution Limits

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

This letter is a request to amend Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant. The proposed changes will revise the Operating License to permit implementation of the Relaxed Axial Offset Control (RAOC) and FQ surveillance methodologies. These methodologies are used to reduce operator action required to maintain conformance with power distribution control Technical Specifications (TS) and to increase the ability to return to power after a plant trip or transient while still maintaining margin to safety limits under all operating conditions.

The changes proposed to TS 3.2.1, Heat Flux Hot Channel Factor FQ(Z); 3.2.3, Axial Flux Difference; and 5.6.5, Core Operating Limits Report (COLR) are being made to adopt the RAOC calculational procedure of the Standard Technical Specifications (STS)

(NUREG-1431, "Standard Westinghouse Technical Specifications Westinghouse Plants"). Changes to TS 3.2.4, Quadrant Power Tilt Ratio (QPTR) are made to provide necessary consistency with the changes made to TS 3.2.1 and TS 3.2.3. Changes to TS 3.3.1 are made to accommodate the change to the RAOC methodologies by revising the form of the f(AI) penalty for overtemperature AT and for overpower AT (initially set to 0).

The adoption of the RAOC and FQ surveillance methodologies are supported by WCAP-10216-P-A, Revision IA (Proprietary), Relaxation of Constant Axial Offset Control FQ -

Surveillance Technical Specification. The NRC has found this WCAP acceptable for referencing in license applications.

Approval of the amendment is requested by April 30, 2006 to provide adequate time to prepare for implementation. Once approved, the amendment will be implemented prior to startup from the fall 2006 refueling outage.

2.0 PROPOSED CHANGE

S This request proposes to modify the Ginna TS by (1) replacing existing specification 3.2.1, Heat Flux Hot Channel (FQ(Z)) methodologies with specification 3.2.1B, Heat Flux Hot Channel (FQ(Z)) from STS that is based on RAOC methodologies; (2) replacing existing specification 3.2.3, Axial Flux Difference (AFD) with specification 3.2.3B, Axial Flux Difference (AFD) from STS that is based on RAOC methodologies and to relocate the requirements for the AFDmonitor alarm from the TS; (3) modifying existing specification 3.2.4, Quadrant Power Tilt Ratio to accommodate the changes made to specifications 3.2.1 and 3.2.3 and to relocate the requirements for the QPTR alarm monitor from the TS; (4) modifying existing specification 3.3.1 to accommodate the use of RAOC methodologies and (5) revise the listing of analysis methodologies contained in TS 5.6.5, Core Operating Limits Report (COLR) to include the references for the RAOC methodology. These changes are necessary to accommodate the planned power uprate and are supported by analyses performed at the planned uprated power level.

Accordingly, the analyses at the planned uprated power level bound operation at the current power level.

An item-by-item list of changes for the proposed changes to specifications 3.2.1 and 3.2.3 is not provided since the existing specifications are being replaced in their entirety with the appropriate STS specifications that are applicable for RAOC methodologies except for the portions of the existing specification 3.2.3 that are associated with the AFD monitor alarm. The existing requirements for the AFD monitor alarm are being relocated to the Technical Requirements Manual (TRM). The STS RAOC specifications 3.2.IB and 3.2.3B are being adopted with no technical deviations from the STS requirements.

Editorial and presentation changes are made to the STS specifications for consistency with the Ginna TS.

The proposed changes revise 3.2.4 Quadrant Power Tilt Ration (QPTR) as follows:

(a) The LCO currently states'The QPTR monitor alarm shall be OPERABLE and QPTR shall be < 1.02"'

The LCO is being revised to state'The QPTR shall be < 1.02" The requirement in the LCO for the QPTR monitor alarm is being removed from the TS and relocated to the Technical Requirements Manual (TRM).

(b) Required Action A. I currently states'Uimit THERMAL POWER to

> 3% below RTP for each 1% of QPTR > 1.00" Required Action A.1 is being revised to state'*Reduce THERMAL POWER > 3%

from RTP for each 1% of QPTR > 1.00" (c) The Completion Time for Required Action A. 1 currently states'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " The Completion Time for Required Action A. 1 is being revised to state'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each QPTR deternination" (d) Required Action A.2 currently states'Perforin SR 3.2.4.2 and limit THERMAL POWER to > 3% below RTP for each 1% of QPTR > 1.00" Required Action A.2 is being revised to state'Determine QPTR" (e) Required Action A.3 currently states'Perform SR 3.2.1.1 and SR 3.2.2.1 Required Action A.3 is being revised to state'Performn SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1 "

(f) The first Completion Time for Required Action A.3 currently states'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months "

-3

The first Completion Time for Required Action A.3 is being revised to state'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A. 1" (g) Required Action A.5 currently states'Normalize excore detector instrumentation to eliminate tilt" Required Action A.5 is being revised to state'Normalize excore detectors to restore QPTR to wvithin limit:"

(h) A second note to Required Action A.5 is being added that states'Required Action A.6 shall be completed whenever Required Action A.5 is performed' (i) The Completion Time for Required Action A.5 currently states'Prior to increasing THERMAL POWER above the limit of Required Actions A. 1 and A.2" The Completion Time for Required Action A.5 is being revised to state'rior to increasing THERMAL POWER above the limit of Required Action A. 1'"

(i) The three existing notes for Required Action A.6 are being replaced with one note that states'Perform Required Action A.6 only after Required Action A.5 is completed" (k) Required Action A.6 currently states'Perform SR 3.2.1.1 and SR 3.2.2.1:"

Required Action A.6 is being revised to state'?erform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2. 1" (I) The first Completion Time for Required Action A.6 currently states"Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching RTP" The first Completion Time for Required Action A.6 is being revised to state"Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions at RTP not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of Required Action A.1" (m)The second Completion Time for Required Action A.6 is being deleted.

(n) Action C is being removed from the TS and relocated to the Technical Requirements Manual (TRM).

(o) SR 3.2.4.1 is being removed from the TS and relocated to the Technical Requirements Manual (TRM). Existing SR 3.2.4.2 is renumbered as SR 3.2.4.1.

(p) The first note to existing SR 3.2.4.2 currently states,'With one power range channel inoperable and THERMAL POWER < 75% RTP, the remaining three power range channels can be used for calculating QPTR" The first note to existing SR 3.2.4.2 is being revised to state'With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER < 75%

RTP, the remaining three power range channels can be used for calculating QPTR" (q) The second note to existing SR 3.2.4.2 currently states'With one power range channel inoperable and THERMAL POWER > 75% RTP, perform SR 3.2.1.2 and SR 3.2.2.2" The second note to existing SR 3.2.4.2 is being revised to state"SR 3.2.4.2 may be performed in lieu of this Surveillance" (r) Existing SR 3.2.4.3 is being removed from the TS.

(s) A new SR 3.2.4.2 with a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency is being added that states'Perform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1" This new SR includes a Note that states'Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP:"

The proposed changes revise 3.3.1 Reactor Trip System (RTS) Instrumentation as follows:

(a) Table 3.3.1-1 function 6, Overpower AT, currently includes SR 3.3.1.3 and SR 3.3.1.6 as surveillance requirements applicable to this function.

SR 3.3.1.3 and SR 3.3.1.6 are being deleted from the list of surveillance requirements applicable to Table 3.3.1-2 function 6, Overpower AT.

(b) Note 1 to Table 3.3.1-1 currently includes the equation for Overtemperature AT. The equation states,

'Overtemperature AT < ATo {Ki + K2 (P-P') - K3 (T-T') [(I +I-xs) / (1+T2 s)] - f(Al))'

The equation for Overtemperature AT is being revised to modify the f(AI) term. The revised equation now states,

'Overtemperature AT < ATo {KI + K2 (P-P') - K3 (T-T') [(1+4ris) / (l+T2 s)]-fj(AI)J' (c) Note 1 to Table 3.3.1-1 currently includes two (2) equations for the f(AI) penalty for Overtemperature AT. The equations currently state, "fRAI) = 0 when qt - qb is < [*]% RTP f(AI) = [*] {(qt - qb) - [*]} when qt - qb is > [*]% RTP' The two (2) equations for f(AI) are being replaced with the following three (3) equations for f1 (AI):

'i(AI) = [*] {[*] - (qt - qb)} when qt - qb < [*]% RTP f1 (AI) =0 % of RTP when [*] % RTP <qt - qb < [*]% RTP f1 (Al) = [*] {(qt - qb) - [*]} when qt - qb > [*]% RTP' (d) Note 2 to Table 3.3.1-1 currently includes the equation for Overpower AT. The equation states,

'Overpower AT = AT. {K4 - K5 (T-T') - K6 [(K3 sT) / (t 3 s+1)] - faI)

The equation for Overpower AT is being revised to modify the f(AI) term. The revised equation now states,

'Overpower AT = AT. {K4 - K5 (T-T') - K6 [(T 3 sT) / (r 3 s+1)]-f 2 (AI F' (e) Note 2 to Table 3.3.1-1 currently includes two (2) equations for the f(AI) penalty for Overpower AT The equations currently state,

'KAI) = [*] when qt-qb < [*] %RTP f(AI) = [*] {(qt-qj}[*]} when qt-qb > [*] %RTP The two (2) equations for f(AI) are being replaced with the following one (1) equation for f2 (AI) f2 (AI)=[*]

The proposed changes revise 5.6.5 Core Operating Limits Report (COLR) as follows:

(t) 5.6.5.b item 3 currently states'WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974. (Methodology for LCO 3.2.3.)

5.6.5.b item 3 is being revised to state'WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification," February 1994. (Methodology for LCO 3.2;1 and LCO 3.2.3.3' Changes to the TS Bases are provided in markup form as Enclosure 4 (for information only). There are no associated Bases for TS 5.6.5. The changes to the Bases include necessary revisions that incorporate the proposed revised Specification verbiage, revise descriptive information to reflect plant specific terminology, wording preferences and design information. The proposed TS changes and associated Bases changes are consistent with STS.

3.0 BACKGROUND

Axial power distribution control at Ginna is currently achieved by the Constant Axial Offset Control (CAOC) methodology. This methodology was developed and described in WCAP-8385 and WCAP-8403. This strategy assures peaking factors and departure from nucleate boiling (DNB) remains'below the accident analysis limits. The CAOC methodology does this by maintaining the axial power distribution within a band of +5%,

-5% Al, for Ginna around a measured target value during normal plant operation, including power changes. By controlling the axial power distribution, the possible skewing of the axial xenon distribution is limited, thus minimizing xenon oscillations and their effects on the power distribution.

Axial Flux Difference (AFD) is a measure of axial power distribution skewing to the top or bottom half of the core. It is sensitive to core-related parameters such as control bank position, core power level, axial burnup, and axial xenon distribution. The limits on AFD assure that the Heat Flux Hot Channel Factor FQ(Z) is not exceeded during either normal operation, or in the event of xenon redistribution following power changes. The AFD limits are used in the nuclear design process and assumed in the safety analyses as a boundary of possible initial condition axial power shapes. Operation outside these AFD limits during Condition I operation (i.e., normal operation) influences the possible power shapes and could result in violations of the kw/fl limit during Condition II transients (i.e.,

Faults of moderate frequency).

An FQ surveillance requirement based tupon the most limiting FQ at each core elevation is presently incorporated into TS 3.2.1, Heat Flux Hot Channel Factor. The CAOC methodology is presently incorporated into TS 3.2.3, Axial Flux Difference. TS 3.2.4, Quadrant Power Tilt Ratio (QPTR) refers to the TS 3.2.1 FQ(Z) surveillance requirements. Application of the RAOC and FQ surveillance methodologies requires the alteration of these TS. A change to TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR), is also required to provide the methodology change.

4.0 TECHNICAL ANALYSIS

The proposed changes to the overtemperature AT equation and the f(AI) equations used in the determination of the overtemperature AT function and the proposed changes to the overpower AT equation and the f(AI) equations used in the determination of the overpower AT function are necessary to accommodate the change to the RAOC methodologies and are based on NRC approved analytical methods (WCAP-8745, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions," March 1977). These changes are consistent with NUREG-1431, Revision 3. NUREG-1431, Revision 3 has been approved for use by the NRC.

The overpower AT function does not require a f(AI) input to the function so the value for f2 (AI) will be zero (0) in the COLR for this parameter. SR 3.3.1.3 and SR 3.3.1.6 are surveillances associated with determination of f(AI). These SRs are no longer necessary for function 6, the overpower AT function, since f2 (AI) will be set to zero (0) for the overpower AT function.

The current TS 3.2.1, Heat Flux Hot Channel Factor and TS 3.2.3, Axial Flux Difference (AFD) are being replaced in their entirety, except for the portions of the existing specification 3.2.3 that are associated with the AFD monitor. The existing requirements for the AFD monitor alarm are being relocated to the Technical Requirements Manual (TRM). The revised specification 3.2.1 and 3.2.3 are consistent with the corresponding specifications in NUREG-1431, Rev. 3 that are appropriate for use with the RAOC methodology. These NUREG-1431, Revision 3, RAOC specific specifications have been approved by the NRC for use in this application. Changes to the TRM are subject to the requirements of 10 CFR 50.59. The TRM revision process and 10 CFR 50.59 requirements provide adequate control for any changes to these AFD monitor alarm requirements.

The current TS 3.2.4, Quadrant Power Tilt (QPTR) is being revised to relocate requirements for the QPTR monitor alarm from the TS to the TRM. TS 3.2.4 is also being revised to accommodate the adoption of the RAOC for TS 3.2.1 and TS 3.2.3. The changes to TS 3.2.4 are consistent with NUREG-143 1, Revision 3 except for SR 3.2.4.2, which will be maintained consistent with the current TS requirements. This amendment request proposes to require the performance of SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1, at a frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , in lieu of the standard SR 3.2.4.2. The SR 3.2.4.2 specified in the NIJREG is based on a capability to use two sets of four thimble locations with quarter core symmetry. The symmetric thimble locations could be used to generate symmetric thimble"tilt'which could then be compared to a reference symmetric thimble"tilt,'from the most recent full core flux map to generate an incore QPTR. Ginna does not have two sets of four thimble locations with quarter core symmetry which are necessary to generate a partial flux map and thus would have to generate a full core flux map. The proposed surveillance and 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency for SR 3.2.4.2 reflects the additional time necessary to generate the full core flux map and is consistent with that specified for the comparable requirement (existing SR 3.2.4.3) in the current TS. The proposed SR 3.2.4.2 is only required to be performed when input from one of more power range neutron flux channel are inoperable with THERMAL POWER > 75% RTP. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. Performing a full core flux map provides an accurate means for ensuring that the core power distribution is consistent with the safety analyses. The proposed Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

NUREG-1431, Revision 3 has been approved for use by the NRC. Changes to the TRM are subject to the requirements of 10 CFR 50.59. The TRM revision process and 10 CFR 50.59 requirements provide adequate control for any changes to these QPTR monitor alarm requirements.

The proposed changes to TS 5.6.5 are being made to be consistent with the RAOC methodology and the STS. A reference to WCAP-10216-P-A, Revision IA is being added because it is applicable to the RAOC methodology. The reference to WCAP-8385 is being deleted because it is applicable to the CAOC methodology, not the RAOC methodology.

The implementation of RAOC and FQ surveillance methodologies have been previously developed and approved by the NRC and documented in WCAP-1 0216-P-A, Rev. IA.

The RAOC strategy was developed to provide wider control bandwidths and more operator freedom than with CAOC. The RAOC methodology provides wider control bands particularly at reduced power by utilizing core margin more effectively. This change provides more operational flexibility in terms of axial power distributions, particularly during power transients such as a return to full power following a power reduction or reactor trip. The wider operating band increases plant availability by permitting increased maneuvering flexibility without a reactor trip or reportable occurrences. The FQ surveillance allows for a more direct surveillance of the elevation-dependent heat flux hot channel factor~and provides margin compared to FQ surveillance requirement that currently exist.

The overall objective of power distribution limits is to provide assurance of fuel integrity during Condition I (Normal Operation) and Condition II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum departure from nucleate boiling ratio (DNBR) in the core greater than or equal to the design DNBR limit during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the loss of coolant accident (LOCA) analyses are met and that the emergency core cooling system (ECCS) acceptance criteria limit of 2200'F is not exceeded with a high probability.

The limits on Axial Flux Difference in a RAOC strategy assure that the FQ(Z) upper bound envelope times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

The limits on heat flux hot channel factor ensure that:

(a) The design limits on peak local power density and minimum DNBR are not exceeded and (b) In the event of a LOCA the peak fuel cladding temperature will not exceed the ECCS acceptance criteria limit of 2200 0 F.

The heat flux hot channel factor is measurable but will normally only be determined periodically as specified in TS 3.2.1 and 3.2.3. This periodic surveillance is sufficient to ensure that the hot channel factor limits are maintained provided:

(a) Control rods in a single group move together with no individual rod insertion differing by more than t12 steps from the group demand position as described in TS 3.1.4,'Rod Group Alignment Limits" (b) Control rod groups are sequenced with overlapping groups as described in TS 3.1.6, "Control Bank Insertion Limits."

(c) The rod insertion limits of TS 3.1.5, "Shutdown Bank Insertion Limit" and TS 3.1.6, "Control Bank Insertion Limits" are maintained.

(d) The axial power distribution, expressed in terms of Axial Flux Difference, is maintained within the limits as described in TS 3.2.3,"Axial Flux Difference (AFD)"

When an FQ measurement is taken, both measurement uncertainty and manufacturing tolerance must be considered. Five percent is the appropriate measurement uncertainty allowance for a full core map taken with the incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

With FQ surveillance, the heat flux hot channel factor FQ(Z) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(Z, BU), to provide assurance that the limit on the heat flux hot channel factor, FQ(Z),

is met. The power factor, W(Z,BU), accounts for the effects of normal operation transients within the AFD band and is determined from expected power control maneuvers over several ranges of burii iP conditions in the core.

An evaluation of the potential impact of the RAOC and FQ(Z) surveillance methodology changes on safety analyses was performed which included:

Non-LOCA Events Non-LOCA Events

LOCA and LOCA-Related Events

Core Design 4.1 Non-LOCA Related Evaluation The effect on the non-LOCA events for a change from CAOC to RAOC is to increase the number of power shapes that must be considered when developing the overtemperature AT and overpower AT setpoint equations. The overtemperature AT setpoint is designed to ensure plant operation within the DNB design basis and hot-leg boiling limit. The overtemperature AT f(AI) function is designed to ensure DNB protection from adverse axial power shapes. The overpower AT trip function is designed to ensure plant operation within the fuel temperature design basis and its required setpoint reduction to maintain FQ(Z) within limits.

Revised overtemperature AT and overpower AT setpoints were developed during the analysis supporting the planned extended power uprate (EPU) to ensure that DNBR design criteria and the fuel temperature design basis will continue to be met. These setpoints are applicable to the change to the RAOC and FQ surveillance methodologies.

The f(AI) function was generated based on the expected axial power shapes from the various Condition I and II events. Because RAOC allows for more severe power shapes to be generated, it was necessary to revise the positive wing and add a negative wing of the overtemperature AT f(AI) penalty to eliminate shapes that may violate the DNB criteria. This wvill have no effect on the Updated Final Safety Analysis Report (UFSAR) transient safety analyses because they do not model the f(AI) term in the overtemperature AT setpoint equation. The f(AI) term accounts for the axial power shape effects on the DNB criteria and independently lowers the overtemperature AT setpoint to ensure a conservative reactor trip. Changes were also made to the overpower AT f(AI) function for the RAOC and FQ surveillances which removed all f(Al) penalties. The penalty will be shown as zero. The implementation of RAOC has been explicitly included in the non-LOCA EPU analyses such that the conclusions presented in the UFSAR remain valid.

The supporting analyses performed for, the EPU bounds operation at the current power level.

4.2 LOCA and LOCA-Related Evaluations The change from CAOC and the current FQ surveillance requirement to the RAOC and FQ(Z) surveillance methodologies has been analyzed in the LOCA safety analyses performed in support of the planned extended power uprate (EPU) program.

The RAOC and FQ(Z) surveillance methodologies do not affect the normal plant operating parameters except for those changed for the EPU (i.e., the axial flux difference, FQ, the overtemperature AT trip setpoint, and the overpower AT trip setpoint parameters),

the safeguards systems actuation, the accident mitigation capabilities important to a LOCA, the assumptions used in the LOCA-related accidents, or create conditions more limiting than those assumed in these analyses.

The main impact of RAOC implementation on the EPU LOCA analyses is the increased range of permissible axial power distributions prior to an event. The projected impact of the RAOC Methodology has been analyzed in both the large-break and small-break LOCA analyses. All related core design parameters will be checked against the LOCA analyses' limits on a cycle-specific basis for reload cycles which utilize the RAOC methodology.

4.3 Core Design Evaluation The changes from CAOC and the current FQ surveillance requirement to RAOC and FQ(Z) surveillance methodologies have been evaluated for impact upon the Ginna core design. Consistent with the approved RAOC methodology, the Condition I axial power shapes were analyzed to demonstrate compliance with the revised LOCA FQ limit. The normal operation axial power shapes were also evaluated relative to the assumed limiting normal operation axial power shape in the analysis of the DNB-limited events which are not terminated by the overtemperature AT reactor trip, e.g. the loss of reactor coolant system flow accident. The Condition II RAOC shapes were analyzed to demonstrate that the fuel melting design criterion was met. In addition, the Condition II axial power distributions were evaluated relative to the axial power distribution assumptions used to generate the DNB core limits. Changes to the axial offset limits and core limits from the EPU analyses were made based on these evaluations. Revised overtemperature AT and overpower AT setpoints were developed during the EPU analysis and applicable to the RAOC and FQ(Z) surveillance methodologies to ensure that DNBR design criteria will continue to be met.

The axial power shapes generated by RAOC were also evaluated in terms of their impact on fuel rod performance. The transient local power increases experienced by the fuel operating within the RAOC Al bands were considered in evaluating the internal pressure of the fuel rods and the cladding transient stress and transient strain. Fuel performance limits were demonstrated to be able to be met under RAOC operation. Compliance with the safety analysis assumptions are performed on a cycle-specific basis during core design analysis.

The use of RAOC and FQ surveillance therefore successfully provides additional operational flexibility to Ginna while still meeting corresponding core design bases and limits.

4.4 Conclusion The technical analysis demonstrates that the implementation of the RAOC and FQ(Z) surveillance methodologies does not affect the normal plant operating parameters except for those changed for the EPU (i.e., the axial flux difference, FQ, the overtemperature AT trip setpoint and the overpower AT trip setpoint parameters), protection system actuation,

Ih I, A: .

the safeguard system actuation, or any other plant capability important to the mitigation of a Non-LOCA or LOCA accident. Potential changes to normal plant operating parameters associated with the EPU will be addressed separately in the EPU submittal.

The supporting analyses performed for the EPU bounds operation at the current power level.

5.0 REGULATORY ANALYSIS

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Ginna LLC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not initiate an accident. Evaluations and analyses of accidents, which are potentially affected by the parameters and assumptions, associated with the RAOC and FQ(Z) methodologies have shown that design standards and applicable safety criteria will continue to be met. The consideration of these changes does not result in a situation where the design, material, or construction standards that were applicable prior to the change are altered.

Therefore, the proposed changes will not result in any additional challenges to plant equipment that could inicrease the probability of any previously evaluated accident.

The proposed changes associated with the RAOC and FQ(Z) methodologies do not affect plant systems such that their function in the control of radiological consequences is adversely affected. The actual plant configurations, performance of systems, or initiating event mechanisms are not being changed as a result of the proposed changes. The design standards and applicable safety criteria limits will continue to be met; therefore, fission barrier integrity is not challenged. The proposed changes associated with the RAOC and FQ(Z) methodologies have been shown not to adversely affect the plant response to postulated accident scenarios.

The proposed changes will therefore not affect the mitigation of the radiological consequences of any accident described in the Updated Final Safety Analysis Report (UFSAR).

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed changes do not challenge the performance or integrity of any safety-related system. The possibility for a new or different type of accident from any accident previously evaluated is not created since the proposed changes do not result in a change to the design basis of any plant structure, system or component. Evaluation of the effects of the proposed changes has shown that design standards and applicable safety criteria continue to be met.

Equipment important to safety will continue to operate as designed and component integrity will not be challenged. The proposed changes do not result in any event previously deemed incredible being made credible. The proposed changes will not result in conditions that are more adverse and will not result in any increase in the challenges to safety systems.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed changes will not involve a significant reduction in a margin of safety.

The proposed changes will assure continued compliance within the acceptance limits previously reviewed and approved by the NRC for RAOC and FQ(Z) methodologies. The appropriate acceptance criteria for the various analyses and evaluations will continue to be met.

The projected impact associated with the implementation of RAOC on peak cladding temperature (PCT) has 'been been incorporated into the LOCA analyses for the planned extended pow'er*'uprate. It has determined that implementation of RAOC at the extended power uprate power level does not result in a significant reduction in a margin of safety. The analysis performed for EPU bounds operation at the current power level.

Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

Based on the above, Ginna LLC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA A review of 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants was conducted to assess the potential impact associated with the proposed changes. Although some UFSAR description of conformance may require a modification, in no case is an exception to any General Design Criterion (GDC) required.

5.2.1 Discussion of Impact The following provides a brief description of GDC 10 and a discussion of the impact on the applicable UFSAR discussion.

GDC 10 Reactor Design state6,' he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design'limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences" UFSAR Discussion/Impact The UFSAR contains analyses of accidents for the Axial Flux Difference parameter. The most important (limiting) Condition II events are the uncontrolled bank withdrawal, cooldown and boration/dilution accidents. The most important (limiting) Condition III and IV events are the loss of flow accident and LOCA, respectively. Calculation of extreme power shapes that affect fuel design limits is performed with approved methods and verified frequently with measurements from the reactors. The conditions under which limiting power shapes are assumed to occur are chosen conservatively with regard to any permissible operating state. To ensure that the axial profile meets the linear heat rate limit and the departure from nucleate boiling (DNB) limit, excore detector signals are used to provide a top to bottom flux difference which is input, through the f(AI), into the overtemperature AT trip setpoint. Nuclear uncertainty margin is applied to calculated peak local power. Such margin is provided for the analysis of normal operating states and for anticipated transients.

This compliance with GDC 10 is not adversely impacted by the proposed changes.

5.2.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in accordance with Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

Ginna LLC has evaluated the proposed changes and determined that:

1. The changes do not involve a significant hazards consideration; and

2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; and

3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.

7.0 REFERENCES

None Fq(Z) (RAOC-W(Z) Methodology) 3.2.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME NOTE B.1 Reduce AFD limits 2 1%for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action B.4 shall be each 1% Fw'(Z) exceeds completed whenever this limit.

Condition Is entered.

AND B. Fw (Z) not within limits. B.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints 2 1%for each 1%

that the maximum allowable power of the AFD limits is reduced.

AND B.3 Reduce Overpower AT trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months setpoints 2 1%for each 1%

that the maximum allowable power of the AFD limits Is reduced.

AND B.4 Perform SR 3.2.1.1 and Prior to increasing SR 3.2.1.2. THERMAL POWER above the maximum allowable power of the AFD limits C. Required Action and C.1 Be In MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

WO)G STS 3.2.1 B-2 Rev. 3.0, 03131/04

Fa(Z) (RAOC-W(Z) Methodology) 3.2.IB3 SURVEILLANCE REQUIREMENTS NOTE During power escalation at the beginning of each cycle, THERMAL POWER may be Increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify Fc(Z) Is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Onc&ewdhin v 23fiours after achieving equilibrium conditions after exceeding, by a 10% RTP, the THERMAL POWER at which Fc (Z) was last verified AND 31 EFPD thereafter WOG STS 3.2.1133 Rev. 3.0. 03/31/04

FI(Z) (RAOC-W(Z) Methodology) 3.2.1 B I SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2 NOTE-If measurements indicate that the maximum over z I FQ (Z) / K(Z)]

has increased since the previous evaluation of FQ (Z):

a. Increasr FQ (Z) by the greater of a factor of f1.02 irby an appropriate factor specified in the COLR and reverify F1 (Z) Is within limits or

b. Repeat SR 3.2.1.2 once per 7 EFPD until either

a. above is met or two successive flux maps Indicate that the maximum overz [Fc(Z) I K(Z)]

has not increased.

Verify Fw (Z) Is within limit. Once after each refueling prior to THERMAL POWER exceed-Ing 75% RTP AND On within 12]ours after achieving equilibrium conditions after exceeding, by z 10% RTP the THERMAL POWER at which Faw(Z) was last verified AND 31 EFPD thereafter WOG STS 3.2.11B4 Rev. 3.0. 03131/04

C<f iGTC c, Etni 4At s - \

o CkIt i A b,i .5 Ftec Z, AFD 3.2.3 3.2 POWERDISTRIBUTION S \ An ckti .1 m>to /

3.2.3 AXIAL FLUX DIFFERENCE (AF ) f Ct sr' /

A rrn.

tiU ; ( ILmDL. ARtu :

a. Shall be maintainelwithin the target band about the target flux difference with3MERMAL POWER 2 90% RTP. The target band Is specified in COLR.

b. May vate outside the target band with THERMAL POWER

< % RTP but 2 50% RTP, provided AFD is within the acceptable operation limits and cumulative penalty deviation time IsS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The acceptable operation limits are

c. May deviate outside the target band wi HERMAL POWER

<50% RTP.

1. The AFD shall be idered outside the target band when the average of four ERABLE excore channels Indicate AFD to be outside the at band. If one excore detector Is out of service, the remaini ree detectors shall be used to derive the average.

2. Pe deviation time shall be accumulated on the basis of a I ute penalty deviation for each 1 minute of power operation with HERMAL POWER 2 50% RTP, and AFD outside the target band.

/3. Penalty deviation time shall be accumulated on the basis of a 0.5 minute penalty deviation for each 1 minute of power operation with THERMAL POWER > 15% RTP and c 50% RTP, and AFD outside the target band.

4. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months of operation be accumulated with AFD outside the target band withgwpenalty deviation time during surveillance of power ra channels In accordance with SR 3.3.1.6. T i

APPLICABILITY: :MODE I with THERMAL POWER > 15% RTP.

R.E. Ginna Nuclear Power Plant 3.2.3-1 Amendment 9

AFD 3.2.3

'LETION TIME lutes lutes iutes

/ -I 4-'

every 15 DS I

R.E. Ginna Nuclear Power Plant 3.2.3-2 Amendments

AFD 3.2.3 iMarrn is lnoperablewhen THERMA KOER N c 9Q% RTP.

V N

2. Assume l6gged values of AFD exislt durling the precedIng 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time interval if a(ftufal AFD //

values are not available.

VerHfVAFD Is within limits and log A rifo each Once within Iour I

i1 OPERABLE excore channel. and every &hour thereafltp

/'

SR 3.2.3.4/ Update target flux differen On within 31 If

/ D after each efueling I

I.

/ 31 EFPD thereafter r-I1~

I*. _ __ _ __.

R.E. Ginna Nuclear Power Plant 3.2.3-3 Amendmentw

AFD 3.2.3 I

R.E. Ginna Nuclear Power Plant 3.2.3-4 Amendments>

AFD jig 3.2.3&j 3.2 POWER DISTRIBUTION LIMITS 3.2.0 AXIAL FLUX DIFFERENCE (AAFD 3 n) WLAxiai )

LCO 3.2.3 The AFD In % flux difference units shall be maintained within the limits specified In the COLR.

NI'-

The AFD shall be considered outside limits when two or more OPERABLE excore channels Indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER 2 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL 30 minutes POWER to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore 7 days channel.

1

QPTR 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR) a Io C I ItI- T fR A`>

LCO 3.2.4 The6-p monitor atitshail be OPERABLE and QPTR shall be S 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS 4`60 . V. '

CONDITION I ( REQUIRED ACTI, ) COMPLETION TIME A. QPTR not within limit A.1 .), THERb1 POWERQ 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sA~J_

2 3% RTP for each 1% of QPTR > 1.00.

f {4- r ea Ch P ANU Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

( ja tr ,1:.:. \

A.3 Perform SR 32.1.11 nd SR 3.2.2.1.

R cALo4 AoAl Once per 7 days thereafter AND A.4 Reevaluate safety analyses Prior to increasing and confirm results remain THERMAL POWER valid for the duration of above the limit of operation under this Required Action A.1 condition.

R.E. Ginna Nuclear Power Plant 3.2.4-1 Arnendment 80

QPTR 3.2.4 CONDITION REQUIRED ACTION ICOMPLETION TIME A.5

_NOTE

/ Perform Required Action A.5 only after Required Av.0 Re si'-I Ad, A.> Action A.4 Is completed.

t\-------___.

Normalize excore etecto Prior to increasing In ( mtai eli ate. THERMAL POWER above the limit of Required AgtionO A.1

_AND r j:e. c~ri r s -1Lct -\

r-rC'}' ee. O;P1 T -k7 V

W88,>

ot W .3 Z^1.i 1 A_~

R.E. Ginna Nuclear Power Plant 3.2.4-2 Amendment 80

QPTR 3.2.4 I

CONDITION I REQUIRED ACTION ICOMPLETION TIME A.6 NOTE -

?t - -or.~~r! ,rC1 14iny requir jto be perfo theaye of AAor A.(- A , (c0 3aociated widi -

IAoperable'QPTR lnstrumenta, tion.

la rCJ

??-1er Ai;

({2. Requireod Action A.6 must be0completed when Req 9 Jfed Action A.5 is

)M AoiL4,- J.,' corpleted and Note 1, apove, does not apply.

1 i Jr;;i b *s <4 R7 p

3. /Only one of he j aA , 6.1 Completion Times.,-/

whichever becomes /r-4H IFr-. m ~

V-r?A D£ W4lt_

applicable first, must be

- met-

. c\ bRev t -Jle

-}l el tjIAcJ ria Perform SR 3.2.1.1' nd SR Withllf24 hours .

3.2.2.1. / after f .TT.l

~3 -J-Z.

@_5-2} 2' Withj48 hours abov94t'elimi'of I i ReqiredA~tions J I .i and A.2 B. Required Action and B.1 ReduceTHERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to S 50% RTP.

Time of Condition A not met.

-=r==:

=:

C. QPTR monitor alarm inoperable,.--

J4I

& ' I C At.

R.E. Ginna Nuclear Power Plant 3.2.4-3 Arnendment 80

OPTR el I-c. o7ff Tat >7 3.2A CONDITION : REQUIRED ACTION ._

C.2P .21.2 and SR Dne-Wifk24 3_.2.2.2. hours and every 24

_- hours thereafter g.

Pa 1 D C#-". e.

-I-a - 4?

SURVEILLANCE REQUIREMENTS*

SURVEILLANCE i FREQUENCY

_ ._., _ _ _. . . .__S (SR 3.2 ,, ._, ,.__

VerifyQPTR monitor-...alarm Is e~

OPERBLE...

- 112.. hours

)

SR 3.:2 42,- NOTE - -

ie.p 9 JAt- Fr o A o.q . "

1.

II) With thanel Inoperable and THERMAL POWE.Ri75% RTP, the remaining ) t;L jSE iFtree power range channels can be used for fclucalculating QPTR.

2. Wt one power range channel inoperable and THERMAL POWER 2 75% RTP, perform SR 2.1.2 and SR 3.2.2.2. - '

Verify QPTR Is within limit by calculation. 7 da~ys~-

1.

.4 D { - NOTE - ,

1. Only required to bwrformed if the OPTR monitor alarm operable.

2. With o ower range channel inoperable and THERMAL POWER < 75% RTP, the remaining ee power range channels can beued for calculating QPTR.

3. With one power range nnel inoperable and ,1.

THERMAL POWE 5% RTP, perform SR 3.2.1.2 and SR .2.2. S-__

Verify QPTR >Within limit by calculation. Once within 24 '

hours and every 24, hours thereafter -'

~ehk AD R.E. Ginna Nuclear Power Plant 3.2.4-4 Amendment 80

Insert A page 3.2.4-4 SR 3.2.4.2 -----------------------NOTE ------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP.

Perform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

.1.

RTS InstrumentatIon I 3.3.1 Tble 3.3.1-1 Reactor Tirp System Instrumentation APPLUCABLE MODES OR LIMITIN OTHER SAFETY SPECIFIED RECURED SURVELLANCE SYSEM FUNCTION CONDmONS CRAHNELS CONDITIONS RECOIREMENTS SETTNOsO

6. OverpowerAT 1.2 4 D.G SR 3.3.1.1 Referto

/.NI. Note 2

, ,I' I:--

SR 33.1.7 SR 3.3.1.10

7. Pressurzer Pressure

a. Low 4 KL SR 3.3.1.1 k 1791 3 SR 33.1.7 P!"g SR 3.3.1.10

b. Fkh 1.2 3 D.G SR 3.3.1 s 239B.2 SR3.3.1.7 paig SR 3.3.1.10

80. Pressurzer Water 1.2 3 DG SR 3.3.1.1 s 96.47%

Level-High SR 3.3.1.7 SR 3.3.1.1 0

9. ReactorCoolant Fbw-Low

a. SiVle Loop 1th) 3perloop M,O SR 33.1.1 k89.88%

SR 3.3.1.7 SR 3.3.1.10 101 *

b. Two Loops j1 Sperloop K.L SR 3.31.1 k 89.S8%

I.

SR 3.3.1.7 SR 3.3.1.tO

10. ReactorCoolant Pmp (RCP)

Breaker Posiron

a. Skgle Loop thi I perRCP N.O SR 3.3.1.11 NA

b. Two Loops tt) I1 perRCP KL SR 3.3.1.11 NA R.E. Ginna Nuclear Power Plant 3.3.1-11 Arnendmenl

RTS Instrumentation 3.3.1 Table 3.3.1-1 (Note 1)

Overlenperature AT

- NOTE -

The Overtemperature T Functton Limiting Safety System Setting Is defined by:

Overtemperature AT s ATo OKI + K2 (P-P) - K3 (T-T) I(14-rs) 1(1+t2s)J -.4i7 Where:

&TIs measured RCS AT, F.

ATO Is the Indicated AT at RTP, F.

s Is the Laplace transform operator, sec 1.

T Is the measured RCS average temperature, OF.

Tr is the nominal T~v, at RTP, OF.

P Is the measured pressurizer pressure, psig.

P' Isthe nominal RCS operating pressure. psig.

/12) Kt Isthe Overtemperature AT reactor trip setpotnt, r].

K2 Is the Overtemperature AT reactor trip depressurizatlon setpolnt penalty coefficient, tlypsl.

K3 is the Overtemperature AT reactor trip heatup setpotnt penalty coefficient, [I'F.

rt Is the measured lead time constant, [1* seconds.

k -z2 Is the measured lag Uime constant, 1) seconds.

f(AI) Is a function of the Indicated difference between the top and bottom detectors of the Power Range Neutron Flux channuls where q, and qb are the percent power In the top and bottom halves of the core, respectively, and qt + qb Is the total THERMAL POWER In percent RTP.

, when4 T- 7j_ rIRTPQ) _

{(qt - qb)- r() when qt - qbpA rj%RTP These values denoted wlth r[ ere specified In the COLR.

R.E. Ginna Nuclear Power Plant 3.3.1-15 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 (Note 2)

OverpowerAT

-NOTE-The Overpower AT Function Lmilling Safety System Setting Is defined by:

Overpower AT 5AT0 (s4 - K5 (T-T) - Ka [(t3sT) I (i3s+1Hl;)

Where:

AT Is measured RCS AT, ,F.

ATo Is the Indicated AT at RTP. IF.

s Is the Laplace transform operator, sec'l.

T is the measured RCS average temperature, "F.

T` Is the nominal T,,v at RTP. 'F.

K4 Is the Overpower AT reactor trip setpolnt, (n.

Ks Is the Overpower AT reactor trip heatup setpolnt penalty coefficient which Is:

LJŽ7 ryjF for T < T' and;

[* rF for T 2 r.

Ka Is the Overpower AT reactor trip thermal time delay setpolnt penalty which Is:

[JI'F for Increastng Tend; j*'JF for decreasing T.

r3 Is the measured Impulselleg time constant, r] seconds.

f(Al) Is a function of the d d difference between the nd bottom detector6 of the Power Range Neutro x channels where qt angare the percent power In the to bottom halves of re, respectIvely, an b Is the total THERMAL PO percent RTP. J

((t=]when q - qb Is.5 [j% RTPJ

These values denoted with '] are specified In the COLR.

(DJ - - - - - ---------

R.E. Ginna Nuclear Power Plant 3.3.1-16 Amendrnentle8

Reporting Requirements 5.6

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described In the following documents:

1. WCAP-9272-P-A, 'Westinghouse Reload Safety Evaluation Methodology,' July 1985.

(Methodology for LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6. LCO 3.2.1, LCO 3.2.2. LCO 3.2.3. and LCO 3.9.1.)

2. WCAP-13677-P-A. '10 CFR 50.46 Evaluation Model Report:

WCOBRAITRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLOTM Cladding Option,* February 1994.

(Methodology for LCO 3.2.1.)

. 1-4A 3. ornContro!ndloed Following)

Repom Septembir 1974, (Methodology for LCO 3.2.3.)

ti_ C , 4. WCAP-1261 0-P-A, 'VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2.1.)

5. WCAP 11397-P-A, "Revised Thermal Design Procedure,'

April 1989.

(Methodology for LCO 3.4.1 when using RTDP.)

6. WCAP-1 0054-P-A and WCAP-1 0081-A, 'Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7. WCAP-1 0924-P-A, Volume 1, Revision 1, *Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:

Model Description and Validation Responses to NRC Questions,' and Addenda 1,2,3, December 1988.

(Methodology for LCO 3.2.1.)

8. WCAP-10924-P-A, Volume 2, Revision 2, 'Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2:

Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.

(Methodology for LCO 3.2.1.)

9. WCAP-1 0924-P-A, Volume 1, Revision 1, Addendum 4,

'Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revislons,' March 1991.

(Methodology for LCO 32.1.)

R.E. Ginna Nuclear Power Plant 5.6-3 AmendmentA%

1*_

Insert A pane 5.6-3 WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification," February 1994.

ENCLOSURE 2 R.E. Ginna Nuclear Power Plant Proposed Technical Specification Changes (markup)

r1/t C.- +J.l -S Sf F(Z)

_+.% S Pe-e- 3.2.1 I 3.2 3.2.1 POWER DISTRIBUTION LIMITS Heat Flux Hot Channel Factor p0 (Z))

M Qt MO r .4

£ LCO 3.2.1 F0 (Z) shall ithin the limits specified in the COLR.

APPLICABILITY: MO 1. /

I ACTIONS CONDi60N REQUIRED ACTION/ COMPLETION TIME II 1 _, _-

A. F0 (Z At within limit. A.1 Reduce THERb AL "I POWER 115 minutes 2 1% RTPffreach 1%

FQ(Z) ex ds limit.

/ /

I A.2 duce AFD acceptable 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months II

/I I. operation limits 2 1% for each 1% FO(Z) exceeds

/.

/ limit.

AND I

II' 11

/ A.3 Reduce Power Range Neutron Flux-High trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

.#e I.

setpolnts 2 1% for each 1%

7Z II

,I FQ(Z) exceeds limit.

AND j I

/

A.4 Reduce Overpowe and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Overtemperatur AT trip setpoInts kIfor each 1%

FQ(Z) ex ds limit.

A.5 Perform SR 3.2.1.1 or SR Prior to Increasing 3.2.1.2. THERMAL POWER above the limit of Required Action A.1 I £ R.E. Ginna Nuclear Power Plant 3.2.1-1 Arnendmente

I X-C> "atr or 3.2 WER DISTRIBUTION LIMITS 3.2.1t\ Heat Flux Hot Channel Factor (Fo(Zj LCO 3.2. Fo(Z), as approximated by Fc(Z) and Fw (Z). shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME NOTE A.1 Reduce THERMAL 15 minutes after each Required Action A.4 shall be POWER 2 1%RTP for Fac(Z) detemmination completed whenever this each 1% Fc (Z) exceeds Condition is entered. limit.

AND A. Fc (Z) not within limit.

A.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each Neutron Flux - High trip Fa (Z) determination setpolnts k 1%for each 1% F09(Z) exceeds limit.

AND A.3 Reduce Overpower AT trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each setpoints 2 1%for each Fac(Z) determination 1% Fc(Z) exceedslimit.

AND AA Perform SR 3.2.1.1 and Prior to increasing SR 3.2.1.2. THERMAL POWER above the limit of Required Action A.1 L3 2He 1

F0 (Z) 3.2.1 "N

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time not met.

B.1 eIn MODE 2.

I 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months I

I SURVEILLANCE RE IREMENTS SURVEILLANCE FREQUENCY 1$

SR 3.2.1.1' Verify measured values of F (Z) Ithin limits Once after each specified In the COLR. refueling prior to THERMAL POWER exceeding 75% 'I RTP I~

v EFPD thereafter

)

SR 3.2.1.2 ------ _1NOTE-----

/nly required to be performed if one powrag

/ cannel is inoperable with THERA WER 2-75%

II 1.

/ RTP.

L Verify measured values of FZ) are within limits specified in the COLR.

Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter

. _ P_

R.E. Ginna Nuclear Power Plant 3.2.1-2 Amendment i

ENCLOSURE 3 R.E. Ginna Nuclear Power Plant Revised Technical Specification Pages (retyped)

FQ(Z) 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQC(Z) and FQW(Z), shall be within the limits I specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

-- __ A.1 ReduceTHERMALPOWER 15 minutes after RuNOTE - A 1% RTP for each 1% each FQC(Z) be completed whenever FQC(Z) exceeds limit. determination this Condition is entered

--_ AND A. FQC(Z) not within limit.

A.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each Neutron Flux - High trip FQC(Z) setpoints 2 1% for each 1% determination FQC(Z) exceeds limit.

AND A.3 Reduce Overpower AT trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each I AND setpoints 2 1% for each 1%

FQC(Z) exceeds limit.

FQC(Z) determination A.4 Perform SR 3.2.1.1 and SR Prior to increasing 3.2.1.2. THERMAL POWER above the limit of Required Action A.1 R.E. Ginna Nuclear Power Plant 3.2.1-1 Amendment

FQ(Z) 3.2.1 CONDITION REQUIRED ACTION COMPLETION TIME

- NOTE--- B.1 Reduce AFD limits 2 1% for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

-NOTE -

Required Action B.4 shall each 1% FQW(Z) exceeds be completed whenever limit.

this Condition is entered.

-- _ AND I B. FQW(Z) not within limits.

B.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints 2 1% for each 1%

that the maximum allowable power of the AFD limits is reduced.

AND B.3 Reduce Overpower AT trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months setpoints 2 1% for each 1%

that the maximum allowable power of the AFD limits is reduced.

AND B.4 Perform SR 3.2.1.1 and SR Prior to increasing 3.2.1.2. THERMAL POWER above the maximum allowable power of the AFD limits C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months I associated Completion Time not met.

R.E. Ginna Nuclear Power Plant 3.2.1-2 Amendment

FO(Z) 3.2.1 SURVEILLANCE REQUIREMENTS

-NOTE-During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is I

obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQC(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75%

RTP AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 2 10% RTP, the THERMAL POWER at which FoC(Z) was last verified AND I 31 EFPD thereafter R.E. Ginna Nuclear Power Plant 3.2.1-3 Amendment

FQ(Z) 3.2.1 SURVEILLANCE FREQUENCY SR 3.2.1.2 -NOTE-If measurements indicate that the maximum over z [FQC(Z) / K(Z) ]

has increased since the previous evaluation of FQC(Z):

a. Increase FQW(Z) by the greater of a factor of 1.02 or by an appropriate factor specified in the COLR and reverify FQW(Z) is within limits or

b. Repeat SR 3.2.1.2 once per 7 EFPD until either

a. above is met or two successive flux maps indicate that the maximum over z [FQC(Z) / K(Z)]

has not increased.

Veri____________is__within__lmit.

I Verify FQW(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75%

RTP AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by

> 10% RTP, the THERMAL POWER at which FQW(Z) was last verified AND I 31 EFPD thereafter R.E. Ginna Nuclear Power Plant 3.2.1-4 Amendment

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits spcified in the COLR.

-NOTE-The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

I APPLICABILITY: MODE 1 with THERMAL POWER 2 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER 30 minutes I to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore 7 days I channel.

R.E. Ginna Nuclear Power Plant 3.2.3-1 Amendment

QPTR 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

I LCO 3.2.4I The QPTR shall be < 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. QPTR not within limit. A.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each I POWER 2 3% from RTP for each 1% of QPTR > 1.00.

QPTR determination AND I A.2 Determine QPTR Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND A.3 Perform SR 3.2.1.1, SR 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 3.2.1.2 and SR 3.2.2.1. achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter AND A.4 Reevaluate safety analyses Prior to increasing and confirm results remain THERMAL POWER valid for the duration of above the limit of operation under this Required Action A.1 condition.

AND R.E. Ginna Nuclear Power Plant 3.2.4-1 Amendment

QPTR 3.2.4 CONDITION E REQUIRED ACTION N COMPLETION TIME A.5

- NOTE -

1. Perform Required Action A.5 only after Required Action A.4 is completed.

2. Required Action A.6 shall be completed whenever Required Action A.5 is performed Normalize excore detectors Prior to increasing to restore QPTR to within THERMAL POWER limit. above the limit of Required Action A.1 AND A.6

-NOTE-Perform Required Action I A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1, SR Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.2.1.2, and SR 3.2.2.1. after achieving equilibrium conditions at RTP not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after increasing THERMAL POWER above the limit of Required Action A.1 l B. Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to S 50% RTP.

Time of Condition A not met.

R.E. Ginna Nuclear Power Plant 3.2.4-2 Amendment

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 -NOTE-

1. With input from one Power Range Neutron Flux I channel inoperable and THERMAL POWER

< 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this I Surveillance.

Verify QPTR is within limit by calculation. 7 days I SR 3.2.4.2 - NOTE -

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER

> 75% RTP.

Pe f o m SR 3. . .1_ R_ _ 2 1 2_ nd S_ 3 2 2_._4 ho r I Perform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months R.E. Ginna Nuclear Power Plant 3.2.4-3 Amendment

RTS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.3.1I The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1-1.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter the Condition Immediately with one channel referenced in Table 3.3.1-1 inoperable. for the channel(s).

OR Two source range channels inoperable.

B. As required by Required B.1 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Action A.1 and referenced OPERABLE status.

by Table 3.3.1-1.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not AND met.

C.2 Initiate action to fully insert 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months all rods.

AND C.3 Place Control Rod Drive 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months System in a condition incapable of rod withdrawal.

R.E. Ginna Nuclear Power Plant 3.3.1-1 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Action A.1 and referenced ------------------

by Table 3.3.1-1. - NOTE -

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months E. As required by Required E.1 Reduce THERMAL POWER 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Action A.1 and referenced to < 5E-11 amps.

by Table 3.3.1-1.

OR E.2 . . . . . . . . .

-NOTE-Required Action E.2 is not applicable when:

a. Two channels are inoperable, or

b. THERMAL POWER is

< 5E-11 amps.

Increase THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to 2 8% RTP.

F. As required by Required F.1 Open RTBs and RTBBs Immediately upon Action A.1 and referenced upon discovery of two discovery of two by Table 3.3.1-1. inoperable channels. inoperable channels AND F.2 Suspend operations Immediately involving positive reactivity additions.

AND F.3 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

R.E. Ginna Nuclear Power Plant 3.3.1-2 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion lime of Condition D, E, or F is not met.

H. As required by Required H.1 Restore at least one 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Action A.1 and referenced channel to OPERABLE discovery of two by Table 3.3.1-1. status upon discovery of two inoperable channels inoperable channels.

AND H.2 Suspend operations Immediately involving positive reactivity additions.

AND H.3 Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

I. Required Action and 1.1 Initiate action to fully insert Immediately associated Completion all rods.

Time of Condition H not met. AND 1.2 Place the Control Rod Drive 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months System in a condition incapable of rod withdrawal.

J. As required by Required J.1 Suspend operations Immediately Action A.1 and referenced involving positive reactivity by Table 3.3.1-1. additions.

AND J.2 Perform SR 3.1.1.1. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter R.E. Ginna Nuclear Power Plant 3.3.1-3 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME K. As required by Required K.1 Action A.1 and referenced ------------------

by Table 3.3.1-1. - NOTE -

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months L. Required Action and L.1 ReduceTHERMALPOWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion to < 8.5% RTP.

Time of Condition K not met.

M. As required by Required M.1 Action A.1 and referenced ------------------

by Table 3.3.1-1. - NOTE -

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months N. As required by Required N.1 Restore channel to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action A.1 and referenced OPERABLE status.

by Table 3.3.1-1.

0. Required Action and 0.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion to < 50% RTP.

Time of Condition M or N not met.

P. As required by Required P.1 Action A.1 and referenced--- --- - ---- ---

by Table 3.3.1-1. - NOTE -

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months R.E. Ginna Nuclear Power Plant 3.3.1-4 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME Q. Required Action and Q.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion to < 50% RTP.

Time of Condition P not met. AND Q.2.1 Verify Steam Dump System 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months is OPERABLE.

OR Q.2.2 Reduce THERMAL POWER 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to < 8% RTP.

R. As required by Required R.1 Action A.1 and referenced ------------------

by Table 3.3.1-1. - NOTE -

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing provided the other train is OPERABLE.

Restore train to OPERABLE 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months status.

S. As required by Required S.1 Verify interlock is in required 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Action A.1 and referenced state for existing plant by Table 3.3.1-1. conditions.

OR S.2 Declare associated RTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Function channel(s) inoperable.

R.E. Ginna Nuclear Power Plant 3.3.1-5 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME T. As required by Required T.1 Action A.1 and referenced----------- - ---

by Table 3.3.1-1. - NOTE -

1. One train may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing, provided the other train is OPERABLE.

2. One RTB may be bypassed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE.

Restore train to OPERABLE 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months status.

U. As required by Required U.1 Restore at least one trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Action A.1 and referenced mechanism to OPERABLE discovery of two by Table 3.3.1-1. status upon discovery of two inoperable trip RTBs with inoperable trip mechanisms mechanisms.

AND U.2 Restore trip mechanism to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

V. Required Action and V.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition R, S, T, or U not met.

W. As required by Required W.1 Restore at least one trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Action A.1 and referenced mechanism to OPERABLE discovery of two by Table 3.3.1-1. status upon discovery of two inoperable trip RTBs with inoperable trip mechanisms mechanisms.

AND R.E. Ginna Nuclear Power Plant 3.3.1-6 Amendment

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME W.2 Restore trip mechanism or 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months train to OPERABLE status.

X. Required Action and X.1 Initiate action to fully insert Immediately associated Completion all rods.

Time of Condition W not met. AND X.2 Place the Control Rod Drive 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months System in a Condition incapable of rod withdrawal.

SURVEILLANCE REQUIREMENTS

- NOTE -

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 - NOTE -

Required to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 2 50% RTP.

Compare results of calorimetric heat balance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months calculation to Nuclear Instrumentation System (NIS) channel output and adjust if calorimetric power is

> 2% higher than indicated NIS power.

SR 3.3.1.3 -NOTE-

1. Required to be performed within 7 days after THERMAL POWER is 2 50% RTP but prior to exceeding 90% RTP following each refueling and if the Surveillance has not been performed within the last 31 EFPD.

2. Performance of SR 3.3.1.6 satisfies this SR.

Compare results of the incore detector measurements 31 effective full to NIS AFD and adjust if absolute difference is 2 3%. power days (EFPD)

R.E. Ginna Nuclear Power Plant 3.3.1-7 Amendment

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.4 Perform TADOT. 31 days on a STAGGERED TEST BASIS SR 3.3.1.5 Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS SR 3.3.1.6 ---- TE-

-NOTE-Not required to be performed until 7 days after THERMAL POWER is 2 50% RTP, but prior to exceeding 90% RTP following each refueling.

Calibrate excore channels to agree with incore 92 EFPD detector measurements.

SR 3.3.1.7 ----

-NOTE-Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 3.

Perform COT. 92 days SR 3.3.1.8 ----

-NOTE-

1. Not required for power range and intermediate range instrumentation until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power < 6% RTP.

2. Not required for source range instrumentation until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power < 5E-11 amps.

Perform COT. 92 days SR 3.3.1.9 - NOTE -

Setpoint verification is not required.

Perform TADOT. 92 days R.E. Ginna Nuclear Power Plant 3.3.1-8 Amendment

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.10 ---- -NOTE-Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.11 Perform TADOT. 24 months SR 3.3.1.12 ----

-NOTE-Setpoint verification is not required.

Perform TADOT. Prior to reactor startup if not performed within previous 31 days SR 3.3.1.13 Perform COT. 24 months R.E. Ginna Nuclear Power Plant 3.3.1-9 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Trip System Instrumentation APPLICABLE MODES OR LIMITING OTHER SAFETY SPECIFIED REQUIRED SURVEILLANCE SYSTEM FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS SETTINGS(a)

1. Manual Reactor Trip 1,2, 2 BC SR 3.3.1.11 NA 3 (b), 4 (b), 5 (b)

2. Power Range Neutron Flux

a. High 1,2 4 D.G SR 3.3.1.1

112.27%

SR 3.3.1.2 RTP SR 3.3.1.7 SR 3.3.1.10

b. Low 1 (c), 2 4 D,G SR 3.3.1.1

29.28%

SR 3.3.1.8 RTP SR 3.3.1.10

3. Intermediate Range 1(c). 2 2 E.G SR 3.3.1.1 (d)

Neutron Flux SR 3.3.1.8 SR 3.3.1.10

4. Source Range 2(e) 2 FG SR 3.3.1.1 (d)

Neutron Flux SR 3.3.1.8 SR 3.3.1.10 3 (b), 4 (b), 5 (b) 2 H.1 SR 3.3.1.1 (d)

SR 3.3.1.7 SR 3.3.1.10 3 (0, 4(1) 5 (f) 1 J SR 3.3.1.1 NA SR 3.3.1.10

5. Overtemperature AT 1.2 4 D,G SR 3.3.1.1 Refer to SR 3.3.1.3 Note 1 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10

6. Overpower AT 1,2 4 D,G SR 3.3.1.1 Refer to l SR 3.3.1.7 Note 2 SR 3.3.1.10 R.E. Ginna Nuclear Power Plant 3.3.1 -10 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Trip System Instrumentation APPLICABLE MODES OR LIMITING OTHER SAFETY SPECIFIED REQUIRED SURVEILLANCE SYSTEM FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS SETTINGS(a)

7. Pressurizer Pressure

a. Low 1(g) 4 KL SR 3.3.1.1 2 1791.3 SR 3.3.1.7 psig SR 3.3.1.10

b. High 1,2 3 DG SR 3.3.1.1 S 2396.2 SR 3.3.1.7 psig SR 3.3.1.10

8. Pressurizer Water 1,2 3 DG SR 3.3.1.1 S 96.47%

Level-High SR 3.3.1.7 SR 3.3.1.10

9. Reactor Coolant Flow-Low

a. Single Loop 1(h) 3 per loop M.O SR 3.3.1.1 2 89.86%

SR 3.3.1.7 SR 3.3.1.10

b. Two Loops 1 (i) 3 per loop KL SR 3.3.1.1 2 89.86%

SR 3.3.1.7 SR 3.3.1.10

10. Reactor Coolant Pump (RCP)

Breaker Position

a. Single Loop 1 (h)

I per RCP N.O SR 3.3.1.11 NA

b. Two Loops 1 per RCP KL SR 3.3.1.11 NA

11. Undervoltage- 1(g) 2 per bus KL SR 3.3.1.9 (d)

Bus 11A and 11B SR 3.3.1.10

12. Underfrequency- 1(g) 2 per bus KL SR 3.3.1.9 2 57.5 HZ Bus 11A and 11B SR 3.3.1.10 R.E. Ginna Nuclear Power Plant 3.3.1-1 1 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Trip System Instrumentation APPLICABLE MODES OR LIMITING OTHER SAFETY SPECIFIED REQUIRED SURVEILLANCE SYSTEM FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS SETTINGS(a)

13. Steam Generator 1,2 3 per SG D,G SR 3.3.1.1 2 13.88%

(SG) Water Level- SR 3.3.1.7 Low Low SR 3.3.1.10

14. Turbine Trip

a. Low Autostop 1 (k)(1) 3 P.Q SR 3.3.1.10 (d)

Oil Pressure SR 3.3.1.12

b. Turbine Stop .1 (k)(1) 2 PQ SR 3.3.1.12 NA Valve Closure

15. Safety Injection (SI) 1,2 2 R,V SR 3.3.1.11 NA Input from Engineered Safety Feature Actuation System (ESFAS)

R.E. Ginna Nuclear Power Plant 3.3.1 -12 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Trip System Instrumentation APPLICABLE MODES OR LIMITING OTHER SAFETY SPECIFIED REQUIRED SURVEILLANCE SYSTEM FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS SETTINGS(a)

16. ReactorTrip System Interlocks

a. Intermediate 2 (e) 2 SNV SR 3.3.1.10 25E-11 Range SR 3.3.1.13 amp Neutron Flux, P-6

b. Low Power 1(g) 4 (power range Sv SR 3.3.1.10

8.0% RTP Reactor Trips only) SR 3.3.1.13 Block, P-7

c. Power Range 1(h) 4 Sv SR 3.3.1.10

49.0%

Neutron Flux, SR 3.3.1.13 RTP P-8

d. Power Range 4 Sv SR 3.3.1.10

50.0%

Neutron Flux, SR 3.3.1.13 RTP P.9 1 (k) 4 Sv SR 3.3.1.10

8.0% RTP SR 3.3.1.13

e. Power Range 1 (c), 2 4 Sv SR 3.3.1.10 2 6.0% RTP Neutron Flux, SR 3.3.1.13 P-10

17. Reactor Trip 1,2 2 trains T.V SR 3.3.1.4 NA Breakers(m) 3 (b), 4 (b), 5 (b) 2 trains WX SR 3.3.1.4 NA

18. Reactor Trip 1,2 1 each per RTB Uv SR 3.3.1.4 NA Breaker 3 (b), 4 (b), 5 (b) 1 each per RTB WX SR 3.3.1.4 NA Undervoltage and Shunt Trip Mechanisms

19. Automatic Trip Logic 1,2 2 trains R,V SR 3.3.1.5 NA 3 (b), 4 (b), 5 (b) 2 trains WX SR 3.3.1.5 NA R.E. Ginna Nuclear Power Plant 3.3.1 -13 Amendment

RTS Instrumentation 3.3.1 (a)

A channel isOPERABLE when both of the following conditions are met:

1. The absolute difference between the as-found Trip Setpoint (TSP) and the previous as-left TSP is within the COT Acceptance Criteria. The COT Acceptance Criteria is defined as:

las-found TSP - previous as-left TSPI

COT uncertainty The COT uncertainty shall not Include the calibration tolerance.

2. The as-left TSP is within the established calibration tolerance band about the nominal TSP. The nominal TSP is the desired setting and shall not exceed the Limiting Safety System Setting (LSSS). The LSSS and the established calibration tolerance band are defined in accordance with the Ginna Instrument Setpoint Methodology. The channel is considered operable even If the as-left TSP is non-conservative with respect to the LSSS provided that the as-left TSP iswithin the established calibration tolerance band.

(b) With Control Rod Drive (CRD) System capable of rod withdrawal or all rods not fully inserted.

(c) THERMAL POWER < 6%RTP.

(d) UFSAR Table 7.2-3.

(e) Both Intermediate Range channels < 5E-11 amps.

(f) With CRD System incapable of withdrawal and all rods fully inserted. Inthis condition, the Source Range Neutron Flux function does not provide a reactor trip, only indication.

(g) THERMAL POWER 2 8.5% RTP.

(h) THERMAL POWER 2 50% RTP.

(I) THERMAL POWER 2 8.5% RTP and Reactor Coolant Flow-Low (Single Loop) trip Function blocked.

() THERMAL POWER 2 8.5% RTP and RCP Breaker Position (Single Loop) trip Function blocked.

(k) THERMAL POWER > 8%RTP, and either no circulating water pump breakers closed, or condenser vacuum

20".

(I) THERMAL POWER 2 50% RTP, 1of 2 circulating water pump breakers closed, and condenser vacuum > 20".

(m) Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

R.E. Ginna Nuclear Power Plant 3.3.1 -14 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 (Note 1)

Overtemperature AT

-NOTE-The Overtemperature AT Function Limiting Safety System Setting is defined by:

Overtemperature AT s AT0 {K1 + K2 (P-P') - K3 (T-T') [(1+Tls) / (1+'r 2s)] - f1 (AI)}

Where:

AT is measured RCS AT, OF.

ATo is the indicated AT at RTP, "F.

s is the Laplace transform operator, sec-1 .

T is the measured RCS average temperature, OF.

T` is the nominal Tavg at RTP, OF.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure, psig.

K1 is the Overtemperature AT reactor trip setpoint, [*].

K2 is the Overtemperature AT reactor trip depressurization setpoint penalty coefficient, [*]/psi.

K3 is the Overtemperature AT reactor trip heatup setpoint penalty coefficient, [*]/OF.

,r is the measured lead time constant, [*] seconds.

x2 is the measured lag time constant, [*] seconds.

f(AI) is a function of the indicated difference between the top and bottom detectors of the Power Range Neutron Flux channels where qt and qb are the percent power in the top and bottom halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

f1(Al) = [*] {[*] - (qt - qb)} when qt - qb * [*]% RTP f1(Al) = 0% of RTP when [*] % RTP < qt - qb * [*]% RTP f1(Al) = [*] {(qt - qb) - []} when qt - qb > [*]% RTP

These values denoted with [*] are specified in the COLR.

R.E. Ginna Nuclear Power Plant 3.3.1 -15 Amendment

RTS Instrumentation 3.3.1 Table 3.3.1-1 (Note 2)

Overpower AT

- NOTE -

The Overpower AT Function Limiting Safety System Setting is defined by:

Overpower AT S ATo {K4 - K5 (T-T') - K6 [(RT3 sT) / (' 3 s+1)] - f2(Al)}

Where:

AT is measured RCS AT, "F.

ATO is the indicated AT at RTP, IF.

s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP, OF.

K4 is the Overpower AT reactor trip setpoint, [*].

K5 is the Overpower AT reactor trip heatup setpoint penalty coefficient which is:

[*]/OF for T < T' and;

[*]/F for T 2 T'.

K6 is the Overpower AT reactor trip thermal time delay setpoint penalty which is:

[*]/OF for increasing T and;

[*]/'F for decreasing T.

T3 is the measured impulse/lag time constant, [*] seconds.

f2 (AI) = [*]

These values denoted with [*] are specified in the COLR.

R.E. Ginna Nuclear Power Plant 3.3.1 -16 Amendment

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Deleted 5.6.2 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 pf each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring activities for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the plant shall be submitted in accordance with 1 0 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Deleted R.E. Ginna Nuclear Power Plant 5.6-1 Amendment

Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

The following administrative requirements apply to the COLR:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

2.1, 'Safety Limits (SLs)";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT (MTC)";

LCO 3.1.5, 'Shutdown Bank Insertion Limit; LCO 3.1.6, 'Control Bank Insertion Limits";

LCO 32.1, "Heat Flux Hot Channel Factor (FO(Z))";

LCO 32.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNAH)";

LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFDr; LCO 3.3.1 "Reactor Protection System (RPS) Instrumentation";

LCO 3A.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boling (DNB) Limits"; and LCO 3.9.1. "Boron Concentration."

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for 2.1, LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2. WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report:

WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLOTM Cladding Option," February 1994.

(Methodology for LCO 3.2.1.)

R.E. Ginna Nuclear Power Plant 5.6-2 Amendment

Reporting Requirements 5.6

3. WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Control / FQ Surveillance Technical Specification," February 1994.

(Methodology for LCO 3.2.1 and LCO 3.2.3.)

4. WCAP-1 2610-P-A, "VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2.1.)

5. WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCO 3.4.1 when using RTDP.)

6. WCAP-1 0054-P-A and WCAP-1 0081-A, 'Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7. WCAP-10924-P-A, Volume 1, Revision 1, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:

Model Description and Validation Responses to NRC Questions," and Addenda 1,2,3, December 1988.

(Methodology for LCO 3.2.1.)

8. WCAP-1 0924-P-A, Volume 2, Revision 2, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2:

Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.

(Methodology for LCO 3.2.1.)

9. WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions," March 1991.

(Methodology for LCO 3.2.1.)

10. WCAP-8745, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions,"

March 1977.

(Methodology for LCO 3.3.1.)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

R.E. Ginna Nuclear Power Plant 5.6-3 Amendment

Reporting Requirements 5.6 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The following administrative requirements apply to the PTLR:

a. RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits"

b. The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP) System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:

LCO 3.4.6, WRCS Loops - MODE 4";

LCO 3.4.7, 'RCS Loops - MODE 5, Loops Filled";

LCO 3.4.10, "Pressurizer Safety Valves"; and LCO 3.4.12, "LTOP System."

c. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC in NRC letter, "R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 2 (TAC No. M96529)," dated November 28,1997. Specifically, the methodology is described in the following documents:

1. Letter from R.C. Mecredy, Rochester Gas and Electric Corporation (RG&E), to Document Control Desk, NRC, Attention: Guy S. Vissing, "Application for Facility Operating License, Revision to Reactor Coolant S ystem (RCS)

Pressure and Temperature Limits Report (PTLR)

Administrative Controls Requirem ents," Attachment VI, September 29, 1997, as supplemented by letter from R.C.

Mecredy, RG&E, to Guy S. Vissing, NRC, "Corrections to Proposed Low Temperature Overpressure Protection System Technical Specification," October 8, 1997.

2. WCAP-1 4040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Sections 1 and 2j January, 1996.

R.E. Ginna Nuclear Power Plant 5.6-4 Amendment

Reporting Requirements 5.6

d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for revisions or supplement thereto.

R.E. Ginna Nuclear Power Plant 5.6-5 Amendment

ENCLOSURE 4 R.E. Ginna Nuclear Power Plant Marked-up Copy of Technical Specification Bases

Reactor Core SLs B 2.1.1 initial conditions of the safety analyses (Ref. 5) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS 21 Figure COLR-5 shows an example of the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is greater than or equal to the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the core exit quality is within the limits defined by the DNBR correlation. Each of the curves of Figure COLR-5 has three distinct slopes. Working from left to right, the first slope ensures that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid such that overtemperature AT indication remains valid. The second slope ensures that the hot leg steam quality remains S 15%. The final slope ensures that DNBR is always;_>Ž 6 j7 lyjff§2*

The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and Al that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

R.E. Ginna Nuclear Power Plant B 2.1 .1-3 Revision

F0 (Z)

B 3.2.1 I

B 3.2 POWER DISTRIBUTIO N LIMITS B 3.2.1 Heat Flux Hot Chani e tor (Fa(Z))

BASES

-/

BACKGR .OUND Te purpose of the limits on the values of F 0(Z) imit the local (i.e.,

pellet) peak power density. The value of F 0 varies along the axial height of the core (Z).

I/' FO(Z) Is defined as the maximum Ifuel rod linear power density

/ divided by the average fuel inear power density, assuming nominal

/ fuel pellet and fuel rod di FO(Z) Is a measure of slons adjusted for uncertainty. Therefore, peak pellet power within the reactor core.

During power raton, the global power distribution Is monitored by LCO 3.2.3,' iAL FLUX DIFFERENCE (AFD),I and LCO 32.4,

'QUAD T POWER TILT RATIO (QPTR)," which are directly and cant ously measured process variables. Therefore, these LCOs serve core limits on a continuous basis.

FQ(Z) Is sensitive to fuel loading pattems, control bank I fuel bumup, and changes In axial power distribution.

/ .. t7 1 ml o

%U *- -SW

SA t5 L m md l pl p r I .

I h , tn ,. 4 ke. tin

S WJI P a Ikj 6r*~ Irjf vU 1 ni lv .4n m.h tO M yl at~S61. I Measurements are generally taken wit e core at or near steady state conditions. With the measured dgedimensional power distributions, It Is possible to determine a mea ed value for Fn(Z). However, because this value represents a ste state condition, it does not include variations In the value (Z), which are present during a nonequilibrium situ n such as load following when the plant changes power level to ch grid demand peaks and valleys.

Core maoring and control under transient conditions (i.e., Condition I events sdescribed In Reference 1) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion, Sequence and Overlap Limits.

/4 ?s 3s .z

\ , ~N C~A ;r ^; wr eo R.E. Ginna Nuclear Power Plant B 3.2,1-1 Revisbon a)

Fa(Z)

B 3.2.1

,- I

/

/APPLICABLE Limits on F0 (Z) preclude core power distributions that violate the SAFETY ANALYSES following fuel design deria:

a. During ass of forced reactor coolant flow accident, there must be I

i at lea 5% probability at the 95% confidence level (the 95195 I I

de ure from nucleate boiling (DNB) criterion) that the hottest fuel in the core does not experience a DNB conditi II b During a large break loss of coolant acciden OCA), peak II cladding temperature (PCT) must not ex d 22000 F (Ref. 2);

c. During an ejected rod accident, the ergy deposition to the fuel will be below 200 cal/gm (Ref. 3 nd I.

I

d. The control rods must be able of shutting down the reactor with a minimum required SH DOWN MARGIN (SDM) with the highest worth control rod stu fully withdrawn (Ref. 4).

Limits on F0(Z) ensu hat the value of the total peaking factor assumed as an initial conditi in the accident analyses remains valid. Other criteria must als met (e.g., maximum cladding oxidation, maximum

.e hydrogen ge coolable geometry, and long term cooling).

iraton, I However, peak cladding temperature is typically most limiting. I The F Z) limits provided In the COLR are based on the limits used in the I LO A analysis. F0(Z) limits assumed in the LOCA analysis are cally I iting relative to (i.e., lower than) the FQ(Z) assumed in s analyses I for other accidents because of the requirements set fo 10 CFR 50.46 (Ref. 2) and ECCS model development In accorda with the required features of the ECCS evaluation models providin 20 CFR 50, Appendix K (Ref. 5). Therefore, this LCO p des conservative limits for other accidents.

FQ(Z) satisfies Criterion 2 of the olicy Statement.

LCO The F 0(Z) shall be ma ned within the limits of the relationships provided In the CO N The F 0(Z) llm define limiting values for core power peaking that precludes ak cladding temperatures above 2200 0F during either a large or all break LOCA (Refs. 6 and 7).

R.E. Ginna Nuclear Power Plant B 3.2.1-2 Revislonw

F0 (Z)

B 3.2.1 This LCO requires operation within the bounds assumed In the safety analyses. Calculaons are performed In the core design process to confirm that th re can be controlled in such a manner during operation that it can within the LOCA F0 (Z) limits. If FQ(Z) cannot be maintad within the LCO limits, reduction of the core power is required.

ating the LCO limits for F0 (Z) may produce unacceptable consequences If a design basis event occurs while Fe(Z) Is outside ts specified limits.

I' APP BILITY The FQ(Z) limits must be maintained w<h MODE I to prevent core power distributions from exceedin , limits assumed In the safety analyses. Applicability In other J DES is not required because there Is neither sufficient stored en In the fuel nor sufficient energy being transferred to the react lant to require a limit on the distribution of core power.

I/

ACT'IONS educing THERMAL POWER by > 1% for each 1% by which F0 (Z) exceeds its limit maintains an acceptable absolute power density. The Completion Time of 15 minutes provides an acceptable time to reduce power In an orderly manner and without allowing the plant to remain in an ar.naccnthabl tnndetinn fwr An smwnn,4aonnAevi ^f fimn 7

When core peaking factors are sufficiently h at LCO 3.2.1 does not permit operation at RTP, the acceptable raon limits for AFD are reduced. The acceptable operation are reduced 1% for each 1% by which F0 (Z) exceeds Its limit. F xample, If the measured Fa(Z) exceeds the limi by 3% an e acceptable operation limits for AFD are

+/- 11% at 90% RTP and 1% at 50% RTP, then the revised AFD Acceptable Operato imits would be +/- 8% at 90% RTP and +/- 28% at 50% RTP. This ures a near constant maximum linear heat rate in I units of kilow Completiq per foot at the acceptable operation limits. The me of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the change In setpoints Is sufficient.

considering the small likelihood of a severe transient In this relatively i,

N-short time period, and the preceding prompt reduction In THERMAL POWER in accordance with Required Action A.1. / ,1 R.E. Ginna Nuclear Power Plant B 32.1-3 Revislong

FQ(Z)

B 3.2.1 A reduction of the Power nge Neutron Flux-High trip setpoints by 2 1%

for each 1% by whic Q(Z) exceeds Its specified limit, Is a conservative '

action for prote c against the consequences of severe transients with ,

unanalyzed er distributions since this trip setpoint helps protect reactor co safety limits. This reduction shall be made as follows, given an F0 limit of 2.32, a measured F0 (Z) of 2.4, and a Power Range Ney n Flux-High setpoint of 108%, the Power Range Neutron Flux-High 7I setpoint must be reduced by at least 3.4% to 104.6%. The Completion l Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient, considering the small likelihood of a severe transient Inthis period, and the preceding prompt reductipn In THERMAL POWER Inaccordance with Required Action A.1.

_A/4 Reduction in the Overpower AT an vertemperature AT trip setpoints by 2 1%for each 1% by which Fq exceeds its limit, is a conservative action for protection again ahe consequences of severe transients with unanalyzed power diqo tions since these trip setpolnts help protect reactor core safety its. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Is sufficient considering the all likelihood of a severe transient In this period, and the precedi prompt reduction InTHERMAL POWER in accordance with Rered Action A.1.

Verification that F0 (Z) has been restored to within Its limit by performing SR 3.2.1.1 or SR 3.2.1.2 prior to increasing THERMAL POWER above the limit Imposed by Required Action A.1 ensures that co nditions during operation at higher power levels are consiste ih safety analyses assumptions.

If the Required Actions of A.1 thr hA.5 cannot be met within their associated Completion Time plant must be placed in a MODE or Condition in which the L90requirements are not applicable. This Is done by placing the GJ-t least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This allowed Completion Time Is reasonable based on operating experience regarding the amount of time It takes to reach MODE 2 from full power operation In an orderly manner and without challenging plant systems.

............ i R.E. Ginna Nuclear Power Plant B 3.2.1-4 Revisionqj

FO(Z)

B 3.2.1

/

II SURVEILLANCE SR 3.2.11J I REQUIREMENTS II I

Verifica n that F 0(Z) is within Its limit involves increasing the measured valu of Fa(Z) to allow for manufacturing tolerance and measurement i

u rtalnties and then making a comparison with the limits. These limits re provided In the COLR. Specifically, the measured value of the Heat I Flux Hot Channel Factor (FMQ) is increased by 3% to account for fuel I manufacturing tolerances and by 5% for flux map measurement I

/ uncertainty for a full core flux map using the mpvable incore detector flux manoina system. This Drocedure Is eauivaiht to Increasino the directly if measured values of Fa(Z) by 8.15% re comparing with LCO limits.

Performing the Surveillance In DE 1 prior to THERMAL POWER exceeding 75% RTP after e refueling ensures that FG(Z) Is within limit when RTP Is achieved a provides confirmation of the nuclear design and the fuel loading p em.

The Frequency I EFPD is adequate for monitoring the change of power distrib n with core bumup because the power distribution changes re ively slowly for this amount of fuel bumup. Accordingly, this Freque Is short enough that the F 0(Z) limit cannot be exceeded for any nificant period of time.

During power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD), and LCO i.,

'QUADRANT POWER TILT RATIO (QPTR)," which are ectty and continuously measured process variables.

With an NIS power range channel Inoperable, TR monitoring for a portion of the reactor core becomes degrad . arge tilts are likely detected with the remaining channels, b he capability for detection of small power tilts In some quadrants Isecreased. Performing SR 3.2.1.2 at a Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provii an accurate alternative means for ensuring that F 0 remains withi mits and the core power distribution is consistent with the safety alyses. A Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months takes Into consideration the rate hich peaking factors are likely to change, and the ime required to bilize the plant and perform a flux map.

This Surveillance is modified by a Note, which states that it Is required only when one power range channel Is Inoperable and the THERMAL POWER is2 75% RTP. of R.E. Ginna Nuclear Power Plant B 3.2.1 -5 Revisiong

FQ(Z)

B 3.2.1 R.E. Ginna Nuclear Power Plant B 3..2.1-6 Revision 21

Fa(Z)

B 3.2 POWER DISTRIBUTION LIMIT B 3.2.1 ot Channel Factor (Fa(Z)

BASES BACKGROUND The purpose of the limits on the values of Fo(Z) is to limit the local (i.e., pellet) peak power density. The value of F0 (Z) varies along the axial height (Z) of the core.

Fo(Z) Is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal pee an roa dzmension s Therefore, FO(Z) Is a measure of the 0a U 1 ^ .- peak fuel pellet power within the reactor core.

1_ During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO(QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis.

Fo(Z) varies with fuel loading patterns, control bank insertion, fuel bumup, and changes in axial power distribution.

Fo(Z) Is measured periodically using the incore detector system. These measurements are generally taken with the core at or near equilibrium A) conditions Using the measured three dimensional power distributions, ItIs possible to derive a measured value for Fa(Z). However, because this value represents an equilibrium condition, itdoes not include the variations In the value of Fo(Z) whlich are present during nonequilibrium situations such as load following or power ascension.

To account for these possible variations, the equilibrium value of Fo(Z) is adjusted as F, (Z) by an elevation dependent factor that accounts for the calculated worst case transient conditions.

Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, an d

_e Havoc 0of

,- ^fens

Insert A page B 3.2.4-1 Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to support flux mapping.

Fo(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES APPLICABLE This LCO precludes core r distributions that violate the following SAFETY fuel design criteria: I ANALYSES

a. During a large rea o of coolant accident (LOCA), the peak cladding temperature ust not exceed 2200"F (Ref. 1), ,r

b. During a loss of forced reactor coolant flow accident, there s a least 95% probability at the 95% confidence level (the 9519wNB) criterion) that the-hotfue-L[ In the cor does not experience a D Bq Ce2;R DNB) condition, c During an ejected rod accident, the energy deposition to the fuel t nt cal/gm (Ref. 2), and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully I withdrawn (Ref. 3).

Umits on Fo(Z) ensure that the value of the Initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the peak cladding temperature Is typically most limiting.

Fo(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the Fa(Z) limit assumed In safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents Fa(Z) satisfies Criterion 2 of 10 CFR 50.3q LCO The Heat Flux Hot Channel Factor, Fa(Z), shall be limited by the following relationships:

Fa(Z)5s(CFQ I P) K(Z) for P>0.5 Fa(Z) S (CFQ / 0.5) K(Z) for P 5 0.5 where: CFQ is the FQ(Z) limit at RTP provided In the COLR, K(Z) Is the normalized FQ(Z) as a function of core height provided in the COLR, and P = THERMAL POWER I RTP B 3.2.IB-2 Rev. 3.0, 03131104 WOG STS WOG STS B 3.2.113-2 Rev. 3.0, 03/31/04

Fo(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES LCO (continued)

For this facility, the actual values of CFO and K(Z) ar in the COLR; however, CFQ is normally a number on the order of and K(Z) is a function that looks like the one provided in Figure B 3.2.11.

For Relaxed Axial Offset Control operation, Fa(Z) Is approximated by F&(Z) and Fw (Z). Thus, both FQ(Z) and Fw (Z) must meet the preceding limits on Fo(Z).

An FQ(Z) evaluation requires obtaining an incore flux map In MODE 1.

From the Incore flux map results we obtain the measured value (F (Z)) of Fc(Z). Then, F&(Z) = Fo(Zx1.0815r'- (fa g) whereE1.081 5Ha factor that accounts for fueI and flux map measurement uncertainty Iem F&(Z) is an excellent approximation for F0 (Z) when the reactor is at the steady state power at which the incore flux map was taken.

A The expression for Fw(Z) Is:

FE Z)-= Fac(Z) W(7 where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) Is included in the COLR. The FQ(Z) is calculated at equilibrium conditions.

The F0 (Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200'F during either a large or small break LOCA.

This LCO requires operation within the bounds assumed in the safety analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that It can stay within the LOCA Fa(Z) limits. If F&(Z) cannot be maintained within the LCO limits, reduction of the core power is required and I FQ (Z) cannot be maintained within the LCO limits, reduction of the AFD limits Is required. Note that sufficient reduction of the AFD limits will also result in a reduction of the core power.

Violating the LCO limits for Fo(Z) produces unacceptable consequences If a design basis event occurs while F0(Z) is outside its specified limis.

B 3.2.IB-3 Rev. 3.0, 03131104 WOG STS WOG STS B 3.2.1 B-3 Rev. 3.0, 03131104

Insert A Dage B 3.2.1-3 Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation.

Fo(Z) (RAOC-W(Z) Methodology) 3.2.1B BASES APPLICABILITY The FO(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed In the safety-analyses.

Applicability In other MODES Is not required because there is either insufficient stored energy In the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.

ACTIONS Ai Reducing THERMAL POWER by 2 1% RTP for each 1% by which F&(Z) exceeds its limit, maintains an acceptable absolute power density. F&(Z) is F.(Z) multiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. FQ (Z) Is the measured value of Fo(Z).

The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time. The maximum allowable power level Initially determined by Required Action A.1 may be affected by subsequent determinations of F&(Z) and would require power reductions within 15 minutes of the Fa(Z) determination, necessary to comply with the decreased maximum allowable power level.

Decreases in Fo(Z) would allow increasing the maximum allowable power level and Increasing power up to this revised limit.

A reduction of the Power Range Neutron Flux - High trip setpoints by Z 1% for each 1% by which FO(Z) exceeds Its limit, Is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Is sufficient considering the small likelihood of a severe transient In this time period and the preceding prompt reduction In THERMAL POWER In accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints Initially determined by Required Action A.2 may be affected by subsequent determinations of F8(Z) and would require Power Range Neutron Flux - High trip set oknt reductions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the Fg(Z) determination, 'necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpolnts. Decreases In F8(Z) would allow Increasing the maximum allowable Power Range Neutron Flux - High trip setpolnts.

B 3.2.IB-4 Rev. 3.0. 03131/04 WOG STS WOG STS B 3.2.1 B- Rev. 3.0. 03131104

Fo(Z) (RAOC-W(Z) Methodology)

B 3.2.18 I BASES ACTIONS (continued)

A.3 Reduction in the Overpower AT trip setpoints (value of K4) by 2 1% for each 1% by which F&(Z) exceeds its limit, Is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER In accordance with Required Action A.1. The maximum allowable Overpower AT trip setpoints Initially determined by Required Action A.3 may be affected by subsequent determinations of Fg(Z) and would rqrOverower AT trip setpoint reductions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the Fo (Z) determination, If necessary to comply with the decreased maximum allowable Overpower AT trip setpoints. Decreases In Fc(Z) would allow increasing the maximum allowable Overpower AT trip setpoints.

A4 Verification that F&(Z) has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.

Condition A Is modified by a Note that requires Required Action A.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to Increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4.

Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure Fa(Z) is properly evaluated prior to Increasing THERMAL POWER.

If It is found that the maximum calculated value of FO(Z) that can occur during normal maneuvers, F.(Z), exceeds is specified limits, there exists A a potential for FQ(z) to become excessively high If a normal operational a te v transient occurs. Reducing the AFD by 2 1% for each 1% by which a A FQ(Z) exceeds Hs limit within the allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , restricts the axial flux distribution such that even If a transient occurred, core peaking factors are not exceededht.jT >

WOG STS B 3.2.1 B-5 Rev. 3.0, 03131/04

Insert A Page B 3.2.1-5 The percent that Fq(Z) exceeds its transient limit is calculated based on the following expression:

maximum ve lover z

[QLCFQ*

[Fc (Z)*W(z)

()

J I *l00 for P >0.5

-Z)K maximum [F (Z) <* W(Z) l100forP<0.5 over z K(z)

L0.5

-F

Fa(Z) (RAOC-W(Z) Methodology)

B 3.2.1 B BASES ACTIONS (continued)

The implicit assumption Is that If W(Z) values were recalculated (consistent with the reduced AFD limits), then Fc(Z) times the recalculated W(Z) values would meet the Fa(Z) limit. Note that complying with this action (of reducing AFD limits) may also result in a power reduction. Hence the need for Required Actions B.2, B.3 and B.4.

B.2 A reduction of the Power Range Neutron Flux-High trip setpoints by > 1%

for each 1%by which the maximum allowable power is reduced, Is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER as a result of reducing AFD limits in accordance with Required Action B.1.

B.3 Reduction in the Overpower AT trip setpoints value of K4 by 2 1%for each 1%by which the maximum allowable power is reduced, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction InTHERMAL POWER as a result of reducing AFD limits In accordance with Required Action B.1.

Verification that FQ(Z) has been restored to within Its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to Increasing THERMAL POWER above the maximum allowable power limit Imposed by Required Action B.1 ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.

B 3.2.16-6 Rev. 3.0. 03131104 WOG STS WOG STS B 3.2.1 B- Rev. 3.0, 03131104

Fo(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES ACTIONS (continued)

Condition B is modified by a Note that requires Required Action B.4 to be performed whenever the Condition is entered. This ensures that SR 3.2.1.1 and SR 3.2.1.2 will be performed prior to increasing THERMAL POWER above the limit of Required Action B.1, even when Cornditiotls exited prior to performing Required Action B.4.

Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure Fo(Z) is properly evaluated prior to increasing THERMAL POWER.

C.1 If Required Actions A.1 through AA or B.1 through B.4 are not met within their associated Completion Times, the plant must be placed In a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant In at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The Note applies REQUIREMENTS during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained.

This allowance Is modified, however, by one of the Frequency conditions that requires verification that FQ(Z) and Fw (Z) are within their specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which they were last verified to be within specified limits.

Because F (Z) and FQ(Z) could not have previously been measured In this reload core, there is a second Frequency condition, applicable only for reload cores, that requires determination of these parameters before exceeding 75% RTP. This ensures that some determination of F (Z) and FQw(Z) are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of FQ(Z) and FQ(Z) following a power increase of more than 10%, ensures that they are verified as soon as RTP (or any other level for extended operation) is achieved. In the absence of these Frequency conditions, it Is possible to B 3.2.16-7 Rev. 3.0. 03/31/04 WOG STS STS B 3.2.1 B-7 Rev. 3.0. 03/31/04

FQ(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES SURVEILLANCE REQUIREMENTS (continued)

Increase power to RTP and operate for 31 days without verification of Fo (Z) and FQ (Z). The Frequency condition is not intended to require verification of these parameters after every 10% Increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which Fa(Z) was last measured.

SR 32.1.1 Verification that FQ(Z) Is within its specified limits Involves increasing F (Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain F&(Z). Specifically, FQ (Z) is the measured valqe of Fa(Z obtaIned from Incore flux map results and F&(Z) = FQ(ZA1.081SlRef. 4). FQ(Z) Is then compared to its specified limits.

The limit with which Fg(Z) is compared varies Inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.

Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the F8(Z) limit is met when RTP Is achieved, because peaking factors generally decrease as power level is increased.

If THERMAL POWER has been increased by 2 10% RTP since the last dete~nination of FQ(Z), another evaluation of this factor Is required

- 1214hours after achieving equilibrium conditions at this higher power level (to ensure that F&(Z) values are being reduced sufficiently with power Increase to stay within the LCO limits).

The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core bumup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS).

B 3.2.18-8 Rev. 3.0, 03131/04 VG STS WOG STS B 3.2.1 B-8 Rev. 3.0, 03/31/04

FO(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.2.1.2 The nuclear design process Includes calculations performed to determine that the core can be operated within the Fo(Z) limits. Because flux maps are taken in steady state conditions, the variations In power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers In normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, FP(Z). by W(Z) gives the maximum FO(Z) calculated to occur In normal operation, FQ(Z).

The limit with which Fw(Z) is compared varies Inversely with power above 50% RTP and directly with the function K(Z) pi ded In the COLR.

The W(Z) curve Is provided Inthe COB P ete core elevations.

Flux map data are typically taken for P ore elevations. FQw(Z) evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% Inclusive and

b. Upper core region, from 85 to 100% inclusive.

The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting In the safety analyses and because of the difficulty of making a precise measurement In these regions.

This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. If Fow(Z) Is evaluated, an evaluation of the expression below Is required to account for any increase to Fo (Z) that may occur and cause the FO(Z) limit to be exceeded before the next required FO(Z) evaluation.

If the two most recent Fo(Z) evaluations show an increase In the expression maximum over z I Fa (Z) I K(Z) 1.It is required to meet the 4Fa(Z) limit with the last Fw'(Z) increased by the greater of a factor of 1.02eor by an appropriate factor specified In the COLR )

B3.2.IB-9 Rev. 3.0, 03131104 WOG STS WOG STS B 3.2.1 B-9 Rev. 3.0, 03/31/04

Fa(Z) (RAOC-W(Z) Methodology)

B 3.2.1B BASES SURVEILLANCE REQUIREMENTS (continued)

--- REVIEWER'S NOTE

/

WCAP-1021P v. 1A, "Relaxation of Colt Axial Offset Control

/and F0 S ance Technical Specific ebruary 1994, ap ate plant specific methooly is to be listed in theR scription in the Adminisratte Controls Section 5.0) dress the methodology used to delive this factor.

or to evaluate FQ(Z) more frequently, each 7 EFPD. These alternative requirements prevent FO(Z) from exceeding its limit for any significant period of time without detection.

Performing the Surveillance In MODE I prior to exceeding 75% RTP ensures that the Fo(Z) limit Is met when RTP is achieved, because peaking factors are generally decreased as power level Is increased.

Fa(Z) Is verified at power levels 2 10% RTP above the THERMAL POWER of its last verification'N1 21f-fours after achieving equilibrium conditions to ensure that Fa(Z) is within its limit at higher power levels.

The Surveillance Frequency of 31 EFPD Is adequate to monitor the change of power distribution with core bumup. The Surveillance may be done more frequently If required by the results of Fa(Z) evaluations.

The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change Is sufficiently slow, when the plant Is operated In accordance with the TS, to preclude adverse peaking factors between 31 day surveillances.

REFERENCES 1. 10 CFR 50.46( 0

2. Ae%;9=; i ip 7 7o-0-n n (A 5 Et1 A

3. I'd 41

4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Chane hactrni@~t Uncertainties," June 1988.

5. WCAP-1 0215-P-A, Rev. IA, "Relaxation of Constant Axial Offset Control (and) F. Surveillance Technical Specification," February 1994.

WOG STS B 3.2.1 B-I 0 Rev. 3.0, 03131104

F0 (Z) (RAOC-W(Z) Methodology)

B 3.2.1B Figure B 3.2.1B-1 (page 1 of 1)

K(Z) - NormalizedZj) as a Function of Core Height WOG STS B 3.2.1 B1 1 Rev. 3.0, 03131104

Insert Apage B 3 2 1-11 1.2 y 1.0

'U C

a Total FO = 2.600 0.

X a6 0 Elevation (fl) 0.0 1.0 12.0 1.0 N

This figure for illustration only, 0 &O Do not use for operation z

0.0 , [ . __

II I 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Elevation (ft)

(31 e S AFD 3.2.3

-Z 'q r r

/B 3.2 POWER DISTRIBUTION LTS i B 3.2.3 AXIAL FLUX DIFFERENII FD) p it AL 6 g1 A' 31 iticl-qd;1-BASES /~ '\ isCae- esy.

BACKGR S6D The purpose of this LCO is to establish lim ithe values of the AFD In order to limit the axial power distr bu skewing to either the top or i bottom of the core. By limiting amount of power distribution skewing, I core peaking factors are c 'istent with the assumptions used in the I safety analyses. LImI Zpower distribution skewing over time also II minimizes the xen distribution skewing, which is a significant factor in II axial power dio uton control. II The operating scheme used to control the axial power distribution, Constant Axial Offset Control (CAOC), involves maintaining the AFD within a tolerance band around a bumup dependent target, known as the

, target flux difference, to minimize the variation of the axial pakirig factor and axial xenon distribution during plant maneuv I i

The target flux difference is determin equilibrium xenon conditions.

The control banks must be positid within the core Inaccordance with I. their Insertion limits and Co Bank D should be Inserted near its i

normal position (I.e., Ž steps withdrawn) for steady state operation at high power levels. e power level should be as near RTP as practical.

The value of arget flux difference obtained under these conditions

/ divided by n of RTP is the target flux difference at RTP for the associa core bumup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RTP value by the appropriate fractional THERMAL POWER level.

II i

- Periodic updating of the target flux difference va ecessary to follow the change of the flux difference at stead conditions with bumup. i i

The Nuclear Enthalpy Rise Hot annel Factor (FNAH) and QUADRANT i POWER TILT RATIO (QPTRrLCOs limit the radial component of the i peaking factors.

I i

II APPLICABLE The AFD Is a measure of axial pow% drution skewing to the top or SAFETY bottom half of the core. The A s sensitive to many core related I

I ANALYSES parameters such as controiank positions, core power level, axial bumup, axial xenondrtbution and, to a lesser extent, reactor coolant I temperature and-1oron concentrations. The allowed range of the AFD Is !

used In the nuclear design process to confirm that operation within these /

limits produces core peaking factors and axial power distributions that )

meet safety analysis requirements.

R.E. Ginna Nuclear Power Plant .B 32.3-1 IRevislong

AFD B 3.2.3 The CAOC methodology (Ref. 1) entails:

a. Establishing an envelop allowed power shapes and power

/ ~densities;:

b. Devising an ratng strategy for the cycle that maximizes pla t-flexibility neuvering) and minimizes axial power shape hnges;

c. De nstrating that this strategy does not result in conditions t violate the envelope of permissible core povwr characteristics; I/nd /

d /Demonstrating that this power distribui ncontrol scheme can be effectively supervised with excore-detectors.

The limits on the AFD ensure t he Heat Flux Hot Channel Factor (Fa(Z)) is not exceeded dunnr either normal operation or Inthe event of xenon redistribution followfng power changes. The limits on the AFD also limit the range of powe(distributions that are assumed as Initial conditions Inanalyzing Condition 2, 3, and 4 events (Ref. 2). This ensures that fuorcladding integrity Is maintained for these postulated ,

accidents. a most important Condition 4 event is the loss of colant accident he most significant Condition 3 event Isthe losspf flow accid The most significant Condition 2 events are uped<ntrolled bank wit, rawal and boration or dilution accidents. Condition 2 accidents, umed to begin from within the AFD limits, amused to confirm the adequacy of Overpower AT and Overtempe tdre AT trip setpoints.

!/ The limits on the AFD satisfy C 7 NRC Policy Statement.

The shape of the po rofile in the axial (i.e., the vertical) direction Is I lamely under the c ol of the operator. throuah either the manual operation of the ntrol banks, or automatic motion of control banks responding t emperature deviations resulting from either manual operation the Chemical and Volume Control System to change boron conce ation, or from power level changes.

nals are available to the operator to help definepYower profile from the Nuclear Instrumentation System (NIS) excord neutron detectors (Ref.

3). Separate signals are taken from the top and bottom excore neutron detectors. The AFD Is defined as the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector in each detector well. For convenience, this flux difference Is converted to provide flux difference units expressed as a percentage and labeled as %Aflux or %Al. /

R.E. Ginna Nuclear Power Plant B 3.2.3-2 RevIsior

AFD B 3.2.3 With THERMAL POWER Z 90Y/yTP (i.e., Part A of this LCO), the AFD must be kept within the targe and about the target flux difference. The target band is provided in e COLR. With the AFD outside the target band with THERMAL PNER 2 90% RTP, the assumptions of the accident analyses mp be violated. With THERMAL POWER c 90%

RTP, the AFD may~e outside the target band provided that the deviation time is restricte5V It Is Intend9 ihat the plant Is operated with the AFD w thin the target band about the target flux difference. However, dcing rapid THERMAL POWiX reductions, control bank motion may use the AFD to deviate outside of the target band at reduced THERY L POWER levels. This deviation does not affect the xenon distri on sufficiently to change the envelope of peaking factors that may b eached on a subsequent return to RTP with the AFD within the targe and, provided the time duration of the deviation is limited. Accordingly while THERMAL POWER Is 2 50%

RTP and < 90% RTP (i.e., Part of this LCO), a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months cumulative penalty deviation time limit, ulative during the preceding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when > 15% RTP, Is allow during which the plant may be operated outside of the target ba but within the acceptable operation limits I

provided In the COL The cumulative penalty time is the sum of penalty times as calculate vyNotes 2 and 3 of this LCO.

For THERMA OWER levels > 15% RTP and < 50% RTP (I.e. Part C of this LCO), viations of the AFD outside of the target band less significant. The reduced penalty deviation time accumulajidl rate reflects this rediced significance. With THERMAL POWER iSA%RTP, AFD Is not a significant parameter In the assumptions usel the safety analysis and, therefore, requires no limits. Because the )ernon distribution produced at THERMAL POWER levels less thin RTP does affect the power distribution as power Is Increased, utanalyzed xenon and power distribution Is prevented by limiting the accumulated penalty deviation time.

The frequency of monitoring th AFD by the Plant Process Computer System (PPCS) Is nominally bnce per minute providing an essentially continuous accumulationot penalty deviation time that allows the-operator to accurately mssess the status of the penalty deviatlottime.

The Inoperability of this monitor requires Independent verification that AFD remains wltrffn limit and that the peaking factors assumed In the accident analyses remain valid.

R.E. Ginna Nuclear Power Plant B 3.2.3-3 Revisiong

AFD B 3.2.3 This LCO Is modified by four Notes. The first Note states the conditio .

necessary for declaring the A outside of the target band. The required>

target band varies with axJumup distribution, which In turn varies with the core average accup lated bumup. The target band defined In the COLR may provideS e target band for the entire cycle or more than one lband, each to bpollowed for a specific range of cycle bLip. The average of thefour OPERABLE excore detectors Is d to determine when AFD is outside the target band. If one exco detector is out of service,.the remaining three detectors are use a derive the average AFD,,The second and third Notes describe ow the cumulative penalty deviation time Is calculated. The second ate states that with THERMAL

,POWER Ž 50% RTP the penalty devin time Is accumulated at the rate of 1 minute for each 1 minute of p er operation with AFD outside the target band. The third Note sta that with THERMAL POWER > 15%

RTP and c 50% RTP the pe deviation time Is accumulated at the rate of 0.5 minutes for ean1 minute of power operation with AFD outside the target ba he cumulative penalty time Is the sum of penalty times from es 2 and 3 of this LCO. The fourth Note addresses AFD outside of th rget band during surveillances. For surveillance of the power ranIchannels performed according to SR 3.3.1.6, deviation outside the et band Is permitted for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months and no penalty deviation time Is ac mulated. Some deviation In the AFD Is required for doing the NIS ca ration with the Incore detector system. This caation Is perted every 92 EFPDs.

!olatingthe LCO on the AFD could produce consequences if a Condition 2, 3, or 4 eve ccepbie rs while the AFD Is outside Its limits.

I/

I APPLIC [TfY AFD requirements are appl e In MODE 1 above 15% RTP. Above 50% RTP, the combinatio of THERMAL POWER and core peaking factors are the core pa eters of primary Importance In safety analyses (Ref. 1). Above 15 TP, this LCO is applicable to ensure that the distributions of xe n are consistent with safety analys ssumptions.

At or below 15 A RTP and for lower operating M he stored energy tES, eg transferred reactor coolant are low Also lohsinalleves i th exorechapls ay recudeobtaining /

vali AF sinalsbelw 1% n P.T~pvalu oftheAFDin these /

R.E. Ginna Nuclear Pczwer Plant B 3.2.3-4 Revision(?

AFD 8 3.2.3 S.s ACTIONS With the AFD outside the targe nd and THERMAL POWER 2 90%

RTP, the assumptions used the accident analyses may be violated with respect to the maximum atIgeneration. Therefore, a Completion Time of 15 minutes is allow to restore the AFD to within the target band because xenon di t utins change little in this relatively short time.

I If Required Acton A.1 is not completed with the required Completion lln)e'of 15 minutes, the axial xenon distribution starts to become skewed.

6ducing THERMAL POWER to < 90% RTP places the corp in a condition that has been analyzed and found to be accepwbre, provided that the AFD Is within the acceptable operation lim P1rovided Inthe COLR.

The allowed Completion Time of 15 minut to reduce THERMAL POWER to < 90% RTP allows for a co lied reduction In power without allowing the plant to remain in an u alyzed condition for an extended period of time.

This Required Action ust be implemented with THERMAL POWER c90% RTP but ; /o RTP Ifeither the cumulative penalty deviation time is > 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months duri the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the AFD Is not within the target band not within the acceptable operation limbs.

With T!3 RMAL POWER < 90% RTP but 2 50% RTP, operation with the AFD/utside the target band Is allowed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the AFD Is within the acceptable operation limits provided In the COLR. With th AFD within these limits, the resulting axial power distribution is acceptable as an Initial condition for accident analyses assum the then existing xenon distributions. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months cumulative pena deviaton time restricts the extent of xenon redistribution. Witho his limitation, unanalyzed xenon axial distributions may result fron.6 different pattern of xenon buildup and decay. Reducing THERMAL POWER to < 50% RTP will put the reactor at a THERMAL POWER level at which the AFD Is not a significant accident analysis parameter.

I If the indicated AFD Is outside the targ'et band and outside the acceptable operation limits provided in the COLR, ti ie peaking factors assumed In accident analysis may be exceeded withi the existing xenon condition.

Any AFD within the target band Is accep,table regardless of its relationship to the acceptable operation limits. The Completion Time of 30 minutes allows for a prompt, yet orderly, reduction In power.

R.E. Glnna Nuclear Power Plant B 3.2.3-5 Revision (;

AFD B 3.2.3 When the AFD monitor alarm Is inoperable and THERMAL POWER Is 2/ 90% RTP. the AFD measure nt determined by the PPCS must be independently monitored to d ect operation outside of the target band and to compute the penalt eviation time at a frequency of every 15 k

minutes to ensure that t plant does not operate In an unanalyzed condition. A Comple n Time of 15 minutes Is adequate to ensure that I the AFD Is within imits at high THERMAL POWER levels and Is consistent with e Completion Time for restoring AFD to within limits tI (Condition A II I

Wh the AFD monitor alarm is Inoperable and THERMAL.0 f0WER is I 9'b% RTP. the AFD measurement determined by thg PPCS must be i ndependently monitored to detect operation outsidoof the target band t and to compute the penalty deviation time at a f oiuency of every hour to ensure that the plant does not operate In an iPanalyzed condition. A Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Is adequate si the AFD may deviate from the target band for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using tp methodology of Notes 2 and 3 of this LCO to calculate the cumulate penalty deviation time before corrective action is required I

EILLANCE SR 3.2.3.1 REM ENTS; This SR Is the rification that the AFD monitor Is OPERABLE. This Is normally a plished by Introducing a signal Into the plant process compute r Iverify control room annunciation of AFD not within the target band. e Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure OPERABILITY of th PD monitor since under normal plant operation, the AFD is r e cted to significantly change.

R 3.2.3.2 The AFD Is monitored on a continuous basis using t Plant Process Computer System (PPCS) that has an AFD mon r alarm. The PPCS determines the 1 minute average of the OPE LE excore detector outputs and provides an alarm message a a main control board annunciator immediately If the average D is outside the target band and then re-alarms when the cumula e penalty deviation Ume reaches 15 minute Intervals within the pre s 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The computer also sends an alarm message whe e cumulative penalty deviation time Is 2 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within the previous hours. This alarm message does not dear until the cumulative nalty deviation time Is < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . /

R.E. Ginna Nuclear Power Plant B 3.2.3-6 Revision lzt

AFD ththe AFDmonitor alarm Inoperable, the AFD measuremen determined by the PPCS must be Independently m ored to detect operation outside of the target band and to cam e the penalty deviation time. During operation at 2 90% RTP, the measurement Is monitored at a Surveillance Frequency 15 minutes to ensure that the AFD Is within Its limits at high THE AL POWER levels. The AFD should be monitored and logge ore frequently during periods of operation for which the pow evel or control bank positions are changing to allow corre y measures when the AFD is more likely to move outside the ta band.

SR 3.2.3.2 Is ified by two Notes. The first Note states that this surveillance only required to be performed when the AFD monitor alarm Is I parable with THERMAL POWER 2 90% RTP. The second Note s es that monitored and logged values of the AFD are as ed to exi ~r the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval in order for the opaef to pute the cumulative penalty deviation time If AFD Yes cannot be

. obtained from the PPCS. InoperabIlity of the alarm Res not necessarily prevent the actual AFD values from being avall (e? . romth computer logs or hand logs). AFD values fo a preceding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months can be obtained from the hourly PPCS print or hand logs.

SR 3.2.3.3/

The AFD is monitored on a nuous basis using the PPCS that has an AFD monitor alarm. The CS determines the 1 minute average of the OPERABLE excore dector outputs and provides an alarm message and a main contra oard annunciator immediately If the average AFD Is outside the targ band and then re-alarms when the cumulative penalty deviation time leaches 15 minute intervals within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The computer also sends an alarm message when the cumulative penalty deviation time Is 2 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This alarm message does not clear until the cumulative penalty deviati n time Is <1 hour within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the AFD monitor alarm Inoperable, the AFD me rement determined by the PPCS must be independently nitored to detect operation outside of the target band and to co ute the penalty deviation time. During operation at -c 90% RTP, b 15% RTP, the AFD

measurement Is monitored at a Surveilla ce Frequency of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure that the AFD Is within its limits"$The Frequency of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Is adequate since the AFD may devoate from the target band for up to I hour using the methodologyof Notes 2 and 3 of this LCO to calculate the cumulative penalty deviation time before corrective action is required.

The AFD should be monitored and logged more frequently in periods of operation for which the power level or control bank positions are changing to allow corrective measures when the AFD Is more likely to move outside the target band.

R.E. Ginna Nuclei ir Power Plant B 3.2.3-7 Revisiong.

AFD B 3.2.3 SR 3.2.3.3 Is modified by two Notes. The first Note states that this surveillance Is only required to be performed when thlKFD monitor alarm Is inoperable with THERMAL POWER < 90 RTP. The second Note states that monitored and logged value the AFD are assumed to exist for the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval invo er for the operator to compute the cumulative penalty deviatiin time if AFD values cannot be:

obtained from the PPCS. Inoperobllty of the alarm does not necessarily prevent the actual AFD values)fJtm being available (e.g., from the computer logs or hand log#) AFD values for the preceding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months can be obtained from the hourly PPCS printouts or hand logs.

SR 3.2.3.4 This Surveillance requires that the target flux difference be updated at a Frequency of 31 effective full power days (EFPD) to account for small changes that may occur in the target flux differences in that period due to There are two methods by which this update can be completede first method requires measuring the target flux difference inacgeeance with SR 32.3.5. This measurement may be obtained usin or excore Instrumentation. The second method involves intr lation between measured and predicted values. The nuclearlesign report provides predicted values for target flux difference arious cycle bumups. The difference between the last measured, ue and the predicted value at

.the same bumup Isapplied to theyrdicted value at the bumup where

'the target flux difference updat! required. This revised predicted value tan then be used to dete e the updated value of the target flux

!iifference.

S'R 3.2.3.5/

Measureme of the target flux difference Isaccomplished by taking aflux map when'the core Is at equilibrium xenon conditions, preferably at power levels with the control banks nearly withdrawn. This flux provides the equilibrium xenon axial power distribution fro.r ch the target value can be determined. The target flux differe varies slowly with core bumup.

A Frequency of once within 31 EFPD after refueling and 92 EFPD thereafter for remeasuring the target flu fferences adjusts the target flux difference to the value measu t steady state conditions. This Is the basis for the CAOC. Remesrement at this Surveillance interval also establishes the AFD ta flux difference values that account for changes Inincore-exco alibrations that may have occurred Inthe Interim.

This SR ism adby a Note that allows the predicted beginning of cycI AFD from the cycle nuclear design to be used to determine the Initial target flux difference after each refuelln R.E. Ginna Nuclear Power Plant B 3.2.3-8 Revision QV L-'

AFD B 3.2.3 1-7, REFERENCES ," WCAP-8403 (nonproprietary), TPowe tribution Control and Load Following Procedures,. We ghouse Electric Corpor fn, September 1974.

Z/

2. American National Sta d, Nuclear Safety Cr aforthe Design of Stationary Pressurized Water Reactor Plants," N18.2-1973. n/

3. UFSAR, Section 7.7.2.6A.4

I R.E. Ginna Nuclear Power Plant -B 3.2.3-9 Revisbon(Ol

I B 3.2 B3.

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD In order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which Is a significant factor in

-axial power distribution control.

1?. 1 J' Av n (RAOC)ls a calculational procedure that defines the allowed operational o Cf s* 1 X, 2 '- ¢ space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidate the conclusions of the accident and transient analyses with regard to fuel cladding integrity.

The AFD Is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels Is outside ts specified limits.

Although I i RAC define"smits that must be met to satisfy safety analyses, typically a ating scheme, Constant Axl Offset Control (CAOC), Is u control axial power distnbuti day to day operation (Ref. 1). C requires that the AFD be aofolled within a narrow tolerle band around a bumup defiant target to minimize the variation of axial peaking facto axial xenon distribution during unit maneuvers.

The CAOC oper pace Is typically smallerrlies within the RAOC operating s . Control within the CAOC 6perating space constrains--

the variaHn of axial xenon distributions and axial power distributions.

lRAOC calculations assume a wide range of xenon distributions and then

confirm that the resulting power distributions satisfy the requirements of ithe accident analyses. .--- -

I o~/ C. cp ifP_

r- Ds1pPaes letl 1.

4 ,0r, s er alt. N 4 0 CA). a Q

AFD (RAOC Methodology)

B 3.2.3B BASES APPLICABLE The AFD is a measure of the axial power distribution skewing to either the SAFETY top or bottom half of the core. The AFD is sensitive to many core related ANALYSES parameters such as control bank positions, core power level, axial bumup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration.

The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions t meet safety analysis requirements.

The RAOC methodology (Ref>l establishes a xenon distribution library with tentatively widq AFD limits. One dimensional axial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysisruiris aTh The limits on the AFD ensure that the Heat Flux lttotinnebFa (Fo(Z)) is not exceeded during either normal oetion or In the event of xenon redistribution following power changes. /lhe limits on the AFD also restrict the range of power distributions that aea used as initial conditions in the analyses of Condition 2, 3, or 4 event* This ensures that the fuel cladding Integrity Is maintained for these postulated accidents. The most Important Condition 4 event is the LOCA. The most important Condition 3 event Is the loss of flow accident. The most important Condition 2 events are

_ ntrolled ban*yithdrawal and boration or dilution accidents.

ondition 2 accidents i ifeto begin from within the AFD limits are

L. A used to confirm the adeiuacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD satisfy Criterion 2 of 10 CFR 50.36*

LCO The shape of the power profile In the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is In response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes.

Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors In each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %Aflux or %Al.

B3.2.3B-2 Rev. 3.0, 03131/04 STS WOG STS B 3.2.313-2 Rev. 3.0, 03131/04

. rlt fi AFD (RAOC Methodology) 1 B 3.2.3B

do 2.

BASES ;

LCO (continued)

The AFD limits are provided In the COLR. Figure B 3.2.4 shows . 4 ZI

%.4- ,

typical RAOC AFD limits. The AFD limits for RAOC do not depend on the .Z l target flux difference. However, the target flux difference may be used to A -....

minimize changes In the axial power distribution. r;~0'!

'- C:M Violating this LCO on the AFD could produce unacceptable l--,o.e \

consequences if a Condition 2, 3, or 4 event occurs while the AFD Is -!at~

outside is specifea imits.

APPLICABILITY The AFD requirements are applicable In MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking , c.i factors are of primary importance In safety analysis. .0 For AFD limits developed using RAOC methodology, the value of the / -mp AFD does not affect the limiting accident consequences with THERMAL j POWER < 50% RTP and for lower operating power MODES. l _

ACTIONS _.7 As an alternative to restoring the AFD to within its specified limits, 7-Required Action A.1 requires a THERMAL POWER reduction to

< 50% RTP. This places the core In a condition for which the value of the AFD Is not Important in the applicable safety analyses. A Completion -i +/- /

Time of 30 minutes Is reasonable, based on operating experience, to I . ;

reach 50% RTP without challenging plant systems. /. ;;

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD, as Indicated by the NIS excore )

channel, Is within Its specified limits. The Surveillance Frequency of i -i 7 days is adequate considering that the AFD Is monitored by a computer ', -

and any deviation from requirements Is alarmed. _ f C REFERENCES 1. ( A03 (nonpropriety neO`5l andLoaad bPowutiob JIollowing ProceSkies,"Westinghouse ElectricCorporation, f~'\eptember)074. /

2. VaIffset Cpntro: F 0 \

onWMsIIere¢9l.,"RelaXatiO>OtCOnStant elanc-nh"nical W F?10217(NP).

June 1983)

3. FSAR, Chapter[i fi7. . .

A P- Ia h6- P-A LA M IJ A Rev. 3.0 03./31/0 WOG STS B 3.2.313-3 Rev. 3.0, 03131104

AFD (RAOC Methodology)

B 3.2.3B I-J I Lj I I=

II--

4.-

0 0

A0 o L

-50 -30 -10 .40 30 50

-40 -20 Oz 20 40 z1 AXIAL FLUX DIFFERENCE (%)

Figure B 3.2. @-1 (page 1 of 1)

AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Function of RATED THERMAL POWER WOG STS B 3.2.313 Rev. 3.0, 03131104

I-l. P ct A- 1! 3.2.3-q I

1 020 -- - - -

_00----=-=-=- = -

(-12, 100)/ (+6, 100)

Unacceptable Unacceptable Operation Operation CE 10

. */ - A cceptab le\.

0 , /Operaio B. 60-- __ -= == --

) 0-40 _I

(-30.50)

-- - 4z==

(+24,50) t 20-l NOTE: This rigure for illustration only. Do not

_ - use for operation.

See COCR W acta operating figure.

50u -4V -;30 -Zu -10 U 10 ZO 30 40 50

Axial Flux Difference (percent AQ

. ., . I. I Fgure B 32.3-1 (page I ol 1)

AXIAL FLUX DIFFERENCE Acceptatle Operation Umits

) 0 as a Function of RATED THERMAL POWER

QPTR B 3.2.4 I B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation. Quadrant Power Tilt Is a core tilt that is measured with the use of the excore power range flux detectors. A core tilt is defined as the ratio of maximum to average quadrant power. The QPTR Is defined as the ratio of the highest average nuclear power In any quadrant to the average nuclear power In the four quadrants. Limiting the QPTR prevents radial xenon oscillations and will Indicate any core asymmetries.

The power density at any point In the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD),' LCO 3.2.4. -QUADRANT POWER TILT RATIO (QPTR),m and LCO 3.1.6, 'Control Bank Insertion Limits,' provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used In the safety analyses.

APPLICABLE Limits on QPTR preclude core power distributions that violate the SAFETY following fuel design criteria:

ANALYSES

a. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hottdffuel rod In the core does not experience a DNB condition;

b. During a large break loss of coolant accident (LOCA), the peak cladding temperature (PCT) must not exceed 22000 F (Ref. 1);

c. During an ejected rod accident, the energy deposition to the fuel will be below 200 caVgm (Ref. 2); and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

R.E. Ginna Nuclear Poewer Plant B 32A-1 Revlslon'9

QPTR B 3.2.4 The LCO limits on the AFD. the QPTR, the Heat Flux Hot Channel Factor (FO(Z)), the Nuclear Enthalpy Rise Hot Channel Factor (FNAH), and Bank Insertion, Sequence and Overlap Limits are established to preclude core power distributions that exceed the safety analyses limits.

The QPTR limits ensure that FN &Hand Fa(Z) remaIn below their limiting values by preventing an undetected change In the gross radial power distribution.

In MODE 1, the FNAH and F0 (Z) limits must be maintained to preclude core power distributions from exceeding design limits assumed In the safety analyses.

The QPTR satisfies Criterion 2 of CFO LCO ahef TR monitor a kshall be OPEE6 E an QPTR shall be (maintai nid at or below the limit of 1.02.

v} 'jeJ ! v° C monitored on automaticbasishusiiibnjt roe;ss

( D nputer System ~PPCS) that has a QPTR o9ar alarm. The PPCS

< ' determines frorrinfe excore detector outp!§e ratio of the highest average nucidar power in any quadra 16the average of nuclear power in the fowfquadrants and providu nalarm message if the QPTR is/

aboveihe 1.02 limit. _

The QPTR limit of 1.02, above which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of SOjcan be tolerated before the_(i margin for uncertainty In FQ(ZJ and P1~AH Is possibly challenged.

cfoWeieUdditionaal QPRof 0.005 is pro-idedrginilh O.7 R.E. Ginna Nuclear Power Plant B 3.2.4-2 Revisioni

QPTR B 3.2A APPLICABILITY The QPTR limit must be maintained in MODE 1 with THERMAL POWER

> 50% RTP to prevent core power distributions from exceeding the design limits assumed In the safety analyses.

Applicability in MODE I s 50% RTP and in other MODES Is not required because there Is neither sufficient stored energy In the fuel nor sufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FNaH and F0 (Z) LCOs still apply below 50% RTP, but allow progressively higher peaking factors as THERMAL POWER decreases below 50%

RTP.

ACTIONS A1 R '__,dtt

With the QPTR exceeding limit, limiing THERMAL POWER to 2 3%

below RTP for each 1% by hich the QPTR exceeds 1.00 Is a conservative tradeoff of to re power with peak linear power. The Completion Time of 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Ilows sufficient time to Identify the cause and correct the tilt. Note that thower reduction itself may cause a.

change In the tilted condition. A further increase Inw-eM R requires a Enaccordanc6fth Required Action :

62 After completion of Required Action A.1, the QPTR alarm may still be In Its alarmed state. As such, any additional changes In the QPTR are detected by requiring a check of the QPTR Cs coordiEc~e with SRT3Ai 2 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. If the QPTR continues to Increase, THERMAL POWER must be limited accordingly. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time Is sufficient because any additional change In QPTR would be relatively slow. , ,(('h& )

Al t f '^j ' -;C()Q .

- The peaking factor s,. Fa(Z) oprimaryImportance In se f I cerP.- cev;w ,-e r ng sution remains consistent with the Initial ' 1?

eI. r U,r-0 i. r Conditions used In the safety analyses. Perimrming SRs on FN a it IVSA L 'FQOZ within the Completion Time of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ansures that thesermy r, ° X iIndicators of power distribution are within their respective limits.A p ( it e @4t; j j 'Completion Time of 24 hourcstakes Into consideration the rate at which gj .1' peaking factors are likely toihange, and the time required to stabilize the

/ plant and perform a flux rn'ap. If these peaking factors are not within their R.E. Ginna Nuclear Power Plant B. 32.4-3 Revision 21i

Insert A page B 3.2.4-3 The maximum allowable power level initially determined by Required Action A. I may be affected by subsequent determinations of QPTR. Increases in the QPTR would require power reductions within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of QPTR determination (completion of applicable surveillance), if necessary to comply with the decreased maximum allowable power level. Decreases in QPTR would allow increasing the maximum allowable power level and increasing power up to this revised limit.

Insert B page B 3.2.4-3 Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to support flux mapping.

QPTR B 3.2.4 limits, the Required Actions of these Surveillances provide an appropriate response for the abnormal condition. If the QPTR remains above Its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate FN ,Hand F0 (Z) with changes In power distribution.

Relatively small changes are expected due to either bumup and xenon redistribution or correction of the cause for exceeding the QPTR limit AA Although FN&H and F0 (Z) are of primary importance as initial conditions In the safety analyses, other changes in the power distribution may occur as the QPTR limit Is exceeded and may have an Impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the OPTR exceeds Its limit, It does not necessarily mean a safety concern exists. It does mean that there Is an indication of a change Inthe gross radial power distribution that requires an Investigation and evaluation that is accomplished by examining the Incore power distributon. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that before Increasing THERMAL POWER to above the limit of Required Action A.1, the reactor core conditions are consistent with the assumptions In the safety analyses.

If thePTRlias exceeded the 1.02 limit and the verification of FN anu F0 (Z) shows that safety requirements are met, the excore detectors are normalized to eliminate thtidicated tlt prior to increasing THERMAL

\ l POWER to above the lmt of Required Actiors A.1 and A.2. This Is done' to detect any subse ient significant cha s In QPTR and to provide a meaningful OPT ala.

Required A ion A.5 Is modlfie; y a Note that states that the Indicated tilt Is not eli nated until after Ih re-evaluation of the safety analysis has

!detemned that core col)dirions at RTP are within the safety analysis assurfptions (I.e., Reqred Action AA). It Is necessary to verify that the cor power distributJd'n is acceptable prior to adjusting the excore detectors to eliminate the Indicated tilt and Increasing power to ensure that the plant is not operating In an unanalyzed condition. This Note is Intended to prevent any ambiguity about the required sequence of actions.

R.E. Ginna Nuclear Power Plant B 32.4-4 Revision gl

QPTR B 3.2.4

-AC.e.' r'

{; VP GAs l9hthe flux tilt is normalized to eliminate the indkcated tilt (i.e., Required Action A.5 Is performed), it Isacceptable to vetum tb full power operation.

However, as an added check that the cdre power;!distribution at RTP Is

consistent with the safety analysis assumptions,.hequired Action A.6

' (C requresverification TM ` that F0 (Z)nd FNHarewlthin their specified limits 3

fC (s)a Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reachingfTP. As an added precaution, if the core

power does not reach wt 24 -htours, but it Increases slowly, then

!the peaking factor su ilannces must be performed within 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months sd[Ei)

(.?h

) 49 ie-ag to p asb Ei i These Completion limes are

)

,ivntended to allow adequate time to increase THERMAL POWER to above

/ the i~mit of Required Actior)A.1 5while not permitting the core to remain with unconfirmed power distributib ns.rtended.periods of time.

71 ~ * $ Re uired Action A.6 ismodified by oterote states that' isnonece X rrformsi - ir6 Actloh 6 1fJUwieauseofthe )

D ,L.J l PTR alari Isassociat Instrumentations#nment The intet of 4A L, his is to clarify the core power djstbution does not havo be Ii; ^ Alarm A) everified if the R Isonly due to the instrumentation (i.e., the' Aa,,t Cxcore dote s) being out of adjupfment and not due to an anomaly:'

p I ,+-; oA Within the re. The sevondIeAtates that the peaking factor F. - > i' D / - i asurveillances after the excore detectors have been U A -normalliebto eliminate the indicated tilt (i.e., Required Action A.5). The 1 - '/ Intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors djusted to eliminate the indicated tilt and the core returned to pow r. he ?dNo1stte tat ly one fthe fdllovtr- pletion Ot llidlmes, whi omrbecomes ap le first, must be rhe Intent of this NoS to clearly Ind hat the first CompUe, Time to be

\c ' app ble Isthe Completon lime which must met to satisfyfoquired;.

If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the plant must be brought to a MODE or condition inwhich the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to S50% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion lime of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

R.E. Ginna Nuclear Power Plant i3 3.2.4-5 Revision 1-l

2PTR B 3.2.4 I

CA. and to.z -

When the QPTR monitor al is inoperable the QPTR must be verified within limits at a frequen of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to ensure hat the plant does not operate In a Rnanalyzed condition. When ERMAL POWER Is 2 75% RTP an ne power range channel Is In rable, QPTR cannot be adequately easured using the excore de tors. In this situation a flux map m be completed to verify that t core power distribution Is consiste the safety analyses. A mpletlon Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequae to detect any relatively slo changes In QPTR, becaus for tho causes of OPT that occur ty kly (e.g., a dropped rod), re tcally o bnormality that prompt a rification of are other Indications I

re power tilt and provides fficient time to stabilize plant and I perform a flux map when cessary. The Complet ime of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is also consistent with th requency of SR 3.24 ith one Inoperable power range chann since these channelsproide input into the QPTRj monitor.

( -.. .._,.

SURVEILLANCE REQUIREMENTS This SR Is the cation that the PTR monitor is OPRAB LE. This Is normally a pished by int cing a signal int e PPCf Verify control annunciatlo QPTR not within li1it. The F quency of 12 hou sufficient to elure OPERABILITy ofthe QPTR monitor since /

u er normal plant peration, OPTR is not expectedrto significantly /

changrr- - ..........-..

ee

. N, 1.Z f This Surveillance verifies that the OPTR, as indicated by the Nuclear.

,~ Instrumentation System (NIS) excore channels, Is within Its limits. The oC. a e t, Frequency of 7 day vhe QPTR alaE ERABLE is<cbptable

44. , (eca(-of the ability that thlsarm can remainl-perable rj, r ithofietecto .

- 6 l sSR 3.5C2.4fs iFi¶ dfied by two Notes. liows QPTR to be

4 I; , .o calculated with three power range channels If THERMAL POWER Is I ^ 75% RTP and one power range channel Is inoperable.J i;isecond

Note states that SR 3.2.1.2 and SR 3.2.2.2 sIoulbdeperformed if

1 r- ,.- THERMAL P0 )Es 2 75% RTP and power range channel is \

inoperable e Intent of this Note oclarify that when one power \

rang annel is Inoperable a HERMAL POWER Is 2 TP. a full j flux map should be ormed to verify the c bwer distribution

. 04 \ stead of using the ee OPERABLE powermge channels to verify 10 \I QPTR. At or alo475% RTP with one powqer range channel inoperable,

\A *t ~ X OPTR monitoring for a portion of the reactor core becomes degraded.

e;. 3. ,^ Large tilts are likely detected with the remaining channels, but the

, aid.0- capability for detection ofsril power tilts in some quadrants Is R.E. Ginna Nuclear Power Plant B 3.2.446 Revision F1

QPTR B 3.2.4 I

. dcased.

_e Performiin full core flux map provds an accurate alterativ s for ensuring thnFH remaln k" core power distribution is consistent with the safty analyses. .

ASR32.4d<.7 fiiiSurveillance verifies that the QPTR, as Ind)cated by the Nuclear I

(Instrumentation System (NIS) excore chan es, Is within Its limits when the QPTR alarm Is Inoperable. The Fr ency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate to Idetect any relatively slow change QPTR, because for those causes of I OPT that occur quickly (e.g., opped rod), there typically are other S-ilndications of abnorrali at prompt a verification of core power tilt.

)C) V SR 3.2.4.3 Is m d by three Notes. The first Note s hat the

/surveillance i ly required to be performed If the monitor alarm Is

{inoperable his surveillance requires a more uent verification that

-1 r the OP is within limit since the monitorp Is Inoperable. The secof Note allows QPTR to be calcuIsfed with three power range

,ch nels if THERMAL POWER is,.75%RTP and one power range channel is Inoperable. The t jrd Note states that SR 3.2.1.2 and SR 3.2.2.2 should be performec if THERMAL POWER is 2 75% RTP and one power range channrel Is Inoperable. The Intent of this Note Is clarify that when one power range channel Is Inoperable and THERMAL POWER is 2 75% RTP, a full core flux map should be perfor i5'erify the core power distribution instead of using the three Q BLE power C Irange channels to verify OPTR. At or above 750 P with one power range channel Inoperable, QPTR monitori core becomes degraded. Large tilts remaining channels, but the cap r a portion of the reactor likely detected with the fordetection of small power tilts In

(

some quadrants Is decreased. Performing a full core flux map provides an accurate alternative means for ensuring that FQ(Z) and FNAH remain l within limits and the core power distribution Is consistent with the safety J analyses. _..

REFERENCES 1. 10 CFR 50A6.

2. UFSAR. Section 15.4.5.

3. Atomic Industrial Forum (AIF) GDC 29, Issued for comment July

10. 1967.

R.E. Ginna Nuclear Power Plant 8 3.2.4-7 Revision

Insert A page B 3.2.4-7 For those causes of quadrant power tilt that occur quickly (e. g., a dropped rod), there typically are other indications of abnormality that prompt a verification of the core power tilt.

Insert B page B 3.2.4-7 This Surveillance is modified by a Note, which states that it is not required until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the input from one or more Power Range Neutron Flux channel is inoperable and the THERMAL POWER is > 75% RTP. With the input from a NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.

When one NIS power range channel input is inoperable and THERMAL POWER is >

75% RTP, a full core flux map should be performed to verify the core power distribution instead of using the three OPERABLE power range channel inputs to verify QPTR by performing SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1, at a Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Performing a full core flux map provides an accurate alternative means for ensuring that FQ(Z) and FN~ remain within limits and the core power distribution is consistent with the safety analyses.

RTS Instrumentation B 3.3.1 account fo4balances in the axialower distributions detectb the NIS upper adIntr power range dfectors.

If l peaks are greater chnedesign limit, as rdcted by we iffrene btwen thefpr and lower poerag

/detectors, the Trip Setp t is reduced in acordance with The Overpower AT trip Function is calculated in two channels for each loop as described in Note 2 of Table 3.3.1-1. A reactor trip occurs If the Overpower AT trip setpoint is reached in two-out-of-four channels. Since the temperature signals are used for other control functions, the actuation logic m ust be able to withstand an input failure to the control system, which may then require the protection function actuation and a single failure in the re maining channels providing the protection function actuation. Section 7.2.5 of Reference 4 discusses control and protection system interactions for this function. Note that th is Function also provides a signal to generate a turbine runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in powe r will normally alleviate the Overpower AT condition and may prevent an unnecessary rea ctor trip.

The LCO requires four channels of the Overpower AT trip Function to be OPERABLE. Note that the Overpower AT trip Function receives input from channels shared with other RTS Functions.

Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overpower AT trip Function must be OPERABLE. These are the only MODES where enough heat is generated in the fuel to be concerned about the heat generation rates and overheating of the fuel. In MODE 3, 4, 5, or 6, this trip Function Is not required to be OPERABLE because the reactor is not operating and there Is insufficient heat production to be concerned about fuel overheating and fuel damage.

7. Pressurizer Pressure The same sensors (PT-429, PT-430, and PT-431) provide Input to the Pressurizer Pressure-High and -Low trips and the Overtemperature AT trip with the exception that the Pressurizer Pressure-Low and Overtemperature AT trips also receive input from PT-449. Since the Pressurizer Pressure channels are also used for other control functions, the actuation logic must be able to withstand an Input failure to the control system, which may then require the protection function actuation, and a single failure In the other channels providing the protection function actuation. Section 7.2.5 of Reference 4 discusses control and protection system interactions for this function.

R.E. Ginna Nuclear Power Plant B 3.3.1-12 Revision 6

RTS Instrumentation B 3.3.1 SR 3.3 1.2 This SR compares the calorimetric heat balance calculation to the NIS Power Range Neutron Flux-High channel output every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is still OPERABLE but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is then declared inoperable.

This SR is modified by a Note which states that this Surveillance is required to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power is 2 50% RTP. At lower power levels, calorimetric data are inaccurate.

The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on plant operating experience, considering instrument reliability and operating history data for instrument drift. Together these factors demonstrate the change in the absolute difference between NIS and heat balance calculated powers rarely exceeds 2% in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

In addition, control room operators periodically monitor redundant indications and alarms to detect deviations in channel outputs.

SR 3.3.1.3 This SR compares the incore system to the NIS channel output every 31 effective full power days (EFPD). If the absolute difference is 2 3%, the NIS channel is still OPERABLE, but must be readjusted. If the NIS channel cannot be properly readjusted, the channel is then declared inoperable. This surveillance is f to ye the f(A) input to the Overtemperature AT Function This SR is modified by two Notes. Note I clarifies that the Surveillance is required to be performed within 7 days after THERMAL POWER is 2 50%

RTP but prior to exceeding 90% RTP following each refueling and if it has not been performed within the last 31 EFPD. Note 2 states that performance of SR 3.3.1.6 satisfies this SR since it is a more comprehensive test.

The Frequency of every 31 EF PD is based on plant operating experience, considering instrument reliabil ity and operating history data for instrument drift. Also, the slow changes in neutron flux during the fuel cycle can be detected during this Interval.

,9R 3.3.1.4 This SR Is the performance of a TADOT every 31 days on a STAGGERED TEST BASIS of the RTB, and the RTB Undervoltage and Shunt Trip Mechanisms. This test shall verify OPERABILITY by actuation of the end devices.

R.E. Ginna Nuclear Power Plant B 3.3.1-37 Revision 6

RTS Instrumentation B 3.3.1 The test shall include separate verification of the undervoltage and shunt trip mechanisms except for the bypass breakers which do not require separate verification since no capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.11. However, the bypass breaker test shall include a local shunt trip. This test must be performed on the bypass breaker prior to placing it

} in service to take the place of a RTB.

The Frequency of every 31 days on a STAGGERED TEST BASIS is based on industry operating experience, considering instrument reliability and operating history data.

SR 3.3.1.5 This SR is the performance of an ACTUATION LOGIC TEST on the RTS Automatic Trip Logic every 31 days on a STAGGERED TEST BASIS.

The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. All possible logic combinations, with and without applicable permissives, are tested for each protection function. The Frequency of every 31 days on a STAGGERED TEST BASIS is based on industry operating experience, considering instrument reliability and operating history data.

SR 3.3.1.6 This SR Is a calibration of the excore channels to the incore channels every 92 EFPD. If the measurements do not agree, the excore channels are still OPERABLE but must be calibrated to agree with the incore detector measurements. If the excore channels cann ot be adjusted, the channels are then declared inoperable. This surveillance is performed to verify the f(Al) input to the Overtem perature AT Function reFpiweZ This SR has been modified by a Note stating that this Surveillance is required to be performed within 7 days after THERMAL POWER is 2 50%

RTP but prior to exceeding 90% RTP following each refueling.

The Frequency of 92 EFPD is adequate based on industry operating experience, considering Instrument reliability and operating history data for instrument drift.

SR 3.3.1.7 This SR is the performance of a COT every 92 days for the following RTS functions:

Power Range Neutron Flux-High;

Source Range Neutron Flux (in MODE 3, 4, or 5 with CRD System capable of rod withdrawal or all rods not fully inserted);

R.E. Ginna Nuclear Power Plant B 3.3.1-38 Revisiong/

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEMS (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the departure from nucleate boiling (DNB) design criterion will be met for each of the transients analyzed.

The design method employed to meet the DNB design criterion for fuel assemblies is the Revised Thermal Design Procedure (RTDP). With the RTDP methodology, uncertainties in plant operating parameters, computer codes and DNB correlation predictions are considered statistically to obtain DNB uncertainty factors. Based on the DNB uncertainty factors, RTDP design limit departure from nucleate boiling ratio (DNBR) values are determined in order to meet the DNB design criterion. Hi aVg . .4E I The RTDP design limit DNBRsfl .244 or the typical and thimble cells for fuel analyses with the WRB-1 correlation.

Additional DNBR margin is maintained by performing the safety analyses to DNBR limits higher than the design limit DNBR values. This margin Sr rse A, o0 n ,/. 0 , between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to 4't ' V.' G F provide DNBR margin for )erating and design flexibility. The safety Ianalysis DNBR value for the typical and thimble cells.

For the WRB-1 correlation, the 95/95 DNBR correlation limit is 1.17. The W-3 DNB correlation is used where the primary DNBR correlation is not applicable. The WRB-1 correlation was developed based on mixing vane data and therefore is only applicable in the heated rod spans above the first mixing vane grid. The W-3 correlation, which does not take credit for mixing vane grids, is used to calculate DNBR values in the heated region below the first mixing vane grid. In addition, the W-3 correlation is applied in the analysis of accident conditions where the system pressure is below the range of the primary correlations. For system pressures in the range of 500 to 1000 psia, the W-3 correlation limit is 1.45. For system pressures greater than 1000 psia, the W-3 correlation limit is 1.30.

R.E. Ginna Nuclear Power Plant B 3.4.1 -1 Revision

ENCLOSURE 5 R.E. Ginna Nuclear Power Plant List of Regulatory Commitments The following table identifies those actions committed to by R.E. Ginna Nuclear Power Plant, LLC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Once approved, the amendment will be Prior to startup from the fall 2006 refueling implemented prior to startup from the fall 2006 outage.

refueling outage.

All related core design parameters will be Prior to startup from the fall 2006 refueling checked against the LOCA analyses limits on a outage.

cycle-specific basis for reload cycles which utilize the RAOC methodology.

v t e Letter Sequence Request
TAC:MC6858, Delete SR Frequencies Based on Inoperable Alarms (Open)
Initiation Request, Request Acceptance, Acceptance Administration Questions 11 August 2005 Answers Results Approval... Other: ML060190605, ML060480244, ML061240036, ML061240563, ML061440090
MONTHYEAR2005-09-19 [Table View]

ML051300330

Text

Subject:

Dear Ms. Skay:

Enclosures:

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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